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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 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

The Transient Reactor Test Facility (TREAT)

TREAT operated from 1959-1994, later refurbished & resumed operation in 2017 to support fuel safety testing Zircaloy-clad graphite/fuel blocks comprise core Virtually any power history possible within ~2000 MJ core transient energy capacity From milliseconds to minutes: Pulses, Ramps, LOCA Fuel motion monitoring system “hodoscope” observes fast neutrons emitted from specimens to track fuel relocation in real time Reactor also can be a neutron source to adjacent radiography facility Experiment vehicle does everything else Safety containment, specimen environment, and instrumentation

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1981

Deposit formation in hydrocarbon rocket fuels: Executive summary

An experimental program was conducted to study deposit formation in hydrocarbon fuels under flow conditions that exist in high-pressure, rocket engine cooling systems. A high pressure fuel coking test apparatus was designed and developed and was used to evaluate thermal decomposition (coking) limits and carbon deposition rates in heated copper tubes for two hydrocarbon rocket fuels, RP-1 and commercial-grade propane. Tests were also conducted using JP-7 and chemically-pure propane as being representative of more refined cuts of the baseline fuels. A parametric evaluation of fuel thermal stability was performed at pressures of 136 atm to 340 atm, bulk fuel velocities in the range 6 to 30 m/sec, and tube wall temperatures in the range 422 to 811K. In addition, the effect of the inside wall material on deposit formation was evaluated in selected tests which were conducted using nickel-plated tubes. The results of the tests indicated that substantial deposit formation occurs with RP-1 fuel at wall temperatures between 600 and 800K, with peak deposit formation occurring near 700K. No improvements were obtained when de-oxygenated JP-7 fuel was substituted for RP-1. The carbon deposition rates for the propane fuels were generally higher than those obtained for either of the kerosene fuels at any given wall temperature. There appeared to be little difference between commercial-grade and chemically-pure propane with regard to type and quantity of deposit. The results of tests conducted with RP-1 indicated that the rate of deposit formation increased slightly with pressure over the range 136 atm to 340 atm. Finally, plating the inside wall of the tubes with nickel was found to significantly reduce carbon deposition rates for RP-1 fuel.

Roback, R.↗
DOE OSTI2022

Milestone 1.2.11: H 2 Production from Surrogate Non-Native Corrosion Plumes on Aluminum 6061-T6 Fuel Cladding Surrogates

Thick, localized, “non-native” corrosion plumes have been identified on Advanced Test Reactor fuel elements, raising concern on their impact on the radiolytic formation of molecular hydrogen gas (H 2 ) from aluminum-clad spent nuclear fuel (ASNF) under proposed extended (> 50 years) dry storage conditions. Here, we report our findings on H 2 generation from the gamma irradiation (up to 52 MGy) of surrogate “non-native” corrosion plume coupons: ambient-temperature-corroded (~350 days in water) aluminum alloy 6061 (AA6061) coupons in helium gas environments with ~0% added relative humidity. Additionally, we provide a comparison of proposed ASNF drying techniques— vacuum drying only, vacuum drying + 100 °C for 4 hr, and vacuum drying + 220 °C for 4 hr—on the yield of H 2 from these surrogate systems. The presented data indicates that similar amounts of H 2 (~2 × 10–3 µmol J–1) are formed from gamma irradiated AA6061 coupons corroded under different temperature regimes, i.e., ambient/350 days vs. 90 C/30 days. These findings validate current, complimentary modelling predictions based on high-temperature-corrosion irradiation data only. Further, the application of a heat treatment procedure (100 and 220 °C), in conjunction with vacuum drying, accelerated the rate at which a steady-state H 2 yield was attained, in comparison to vacuum only, due to the removal of H 2 precursors in the form of adsorbed waters. Interestingly, within the confidence limits of our measurements, negligible difference in total H 2 yield was found between the two investigated heat treatment procedures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
NASA NTRS2019

Infrared Measurements of Forward Heat Conduction during Simulated Microgravity Flame Spread in the Narrow Channel Apparatus

An infrared (IR) camera provides a way of examining temperature trends associated with simulated microgravity flame spread in the Narrow Channel Apparatus (NCA). The IR camera measures the surface temperature of solid poly methyl methacrylate (PMMA) fuel. These tests examine the forward conduction of heat ahead of the flame front in the non-thermally thin fuel.The NCA is a combustion wind tunnel that simulates a microgravity flame spread environment by employing a narrow gap between the fuel and ceiling of the device, limiting the effects of buoyancy. Test conditions of a 5 mm gap, mean opposed flow velocity of 15 cm/s, and fuel thickness of 3 mm are used.PMMA is selected as the fuel due to repeatability of test results, ease of computational modeling, and known combustion mechanics. Using specific lens and bandpass filter combinations the PMMA can be imaged as effectively opaque. The spectral emissivity for PMMA was calculated and incorporated into the calibration of the camera.Surface temperatures from the IR camera are compared to results from thermocouples embedded in the surface of the fuel. The IR camera results show that nontrivial forward conduction occurs during tests, and therefore must be included in computational models of the process.

Berry, Michael↗
NASA NTRS1985

Fuel containment and damage tolerance in large composite primary aircraft structures. Phase 2: Testing

Technical issues associated with fuel containment and damage tolerance of composite wing structures for transport aircraft were investigated. Material evaluation tests were conducted on two toughened resin composites: Celion/HX1504 and Celion/5245. These consisted of impact, tension, compression, edge delamination, and double cantilever beam tests. Another test series was conducted on graphite/epoxy box beams simulating a wing cover to spar cap joint configuration of a pressurized fuel tank. These tests evaluated the effectiveness of sealing methods with various fastener types and spacings under fatigue loading and with pressurized fuel. Another test series evaluated the ability of the selected coatings, film, and materials to prevent fuel leakage through 32-ply AS4/2220-1 laminates at various impact energy levels. To verify the structural integrity of the technology demonstration article structural details, tests were conducted on blade stiffened panels and sections. Compression tests were performed on undamaged and impacted stiffened AS4/2220-1 panels and smaller element tests to evaluate stiffener pull-off, side load and failsafe properties. Compression tests were also performed on panels subjected to Zone 2 lightning strikes. All of these data were integrated into a demonstration article representing a moderately loaded area of a transport wing. This test combined lightning strike, pressurized fuel, impact, impact repair, fatigue and residual strength.

Sandifer, J. P.↗
DOE OSTI2019

Fuel Loading Optimization and Planning for the Versatile Test Reactor

The Versatile Test Reactor (VTR) mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. In service of this duty, the goal of the design is to maximize the peak fast flux and the experimental volume exposed to a fast flux above 4.0 × 1015n/cms s. Delivering the desired neutron fluence for each experiment requires the capability to optimize the fuel loading to achieve a specified cycle length and flux, while also satisfying other criteria such as thermal hydraulic design constraints. To achieve the desired conditions and cycle length, several measures may be taken, including optimization of the fuel loading pattern, moving partially-burned assemblies back and forth between the core and the in-vessel storage, and potentially shuffling fuel within the core. Optimizing in a "brute-force" manner by solving a wholecore transport equation for hundreds of configurations would be inefficient. A sufficiently refined DIF3D-VARIANT solution for the whole core requires on the order of 1 hour. Instead, we seek an approximate evaluation of the parameters for many variations on a base core configuration without directly solving a diffusion or transport equation for each.

fuel loading optimization↗
NASA NTRS1973

Flowing gas, non-nuclear experiments on the gas core reactor

Variations in cavity wall and injection configurations of the gas core reactor were aimed at establishing flow patterns that give a maximum of the nuclear criticality eigenvalue. Correlation with the nuclear effect was made using multigroup diffusion theory normalized by previous benchmark critical experiments. Air was used to simulate the hydrogen propellant in the flow tests, and smoked air, argon, or Freon to simulate the central nuclear fuel gas. Tests were run both in the down-firing and upfiring directions. Results showed that acceptable flow patterns with volume fraction for the simulated nuclear fuel gas and high flow rate ratios of propellant to fuel can be obtained. Using a point injector for the fuel, good flow patterns are obtained by directing the outer gas at high velocity long the cavity wall, using louvered injection schemes. Recirculation patterns were needed to stabilize the heavy central gas when different gases are used.

Kunze, J. F.↗
DOE OSTI2023

Auto Stop-Start Fuel Consumption Benefits

With increasingly stringent regulations mandating the improvement of vehicle fuel economy, automotive manufacturers face growing pressure to develop and implement technologies that improve overall system efficiency. One such technology is an automatic (auto) stop-start feature. Auto stop-start reduces idle time and reduces fuel use by temporarily shutting the engine off when the vehicle comes to a stop and automatically re-starting it when the brake is released, or the accelerator is pressed. As mandated by the U.S. Congress, the U.S. Environmental Protection Agency (EPA) is required to keep the public informed about fuel saving practices. This is done, in partnership with the U.S. Department of Energy (DOE), through the fueleconomy.gov website. The “Fuel-Saving Technologies” and “Gas Mileage Tips” sections of the website are focused on helping the public make informed purchasing decisions and encouraging fuel-saving driving habits. Here, in order to provide users with accurate information about the auto stop-start feature, experiments were conducted to determine its fuel economy effect. Four vehicles were tested both with and without the feature enabled under three test cycles: the Federal Test Procedure (FTP) city fuel economy test, the US06 high acceleration aggressive driving schedule that is often identified as the “Supplemental FTP” driving schedule, and the EPA New York City Cycle (NYCC). The results were compared to measure the fuel economy and consumption effects of using the auto stop-start feature. It was found that the fuel economy improvement varied significantly between drive cycles depending on the amount and percentage of idle time during the test. The largest fuel economy improvements were 7.27% and 26.4% for the FTP and NYCC, respectively.

33 ADVANCED PROPULSION SYSTEMS↗
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↗
NASA NTRS1982

Evaluation of candidate stirling engine heater tube alloys at 820 deg and 860 deg C

Seven commercial alloys were evaluated in Stirling simulator materials rigs. Five iron base alloys (N-155, A-286, Incoloy 800, 19-9DL, and 316 stainless steel), one nickel base alloy (Inconel 718), and one cobalt base alloy (HS-188) were tested in the form of thin wall tubing in a diesel fuel fired test rig. Tubes filled with hydrogen or helium at gas pressure of 21.6 MPa and temperatures of 820 and 860 C were endurance tested for 1000 and 535 hours, respectively. Results showed that under these conditions hydrogen permeated rapidly through the tube walls, thus requiring refilling during each five hour cycle. Helium was readily contained, exhibiting no measurable loss by permeation. Helium filled tubes tested at 860 C all exhibited creep-rupture failures within the 535 hour endurance test. Subsequent tensile test evaluation after removal from the rig indicated reduced room temperature ductility for some hydrogen-filled tubes compared to helium-filled tubes, suggesting possible hydrogen embrittlement in these alloys.

Misencik, J. A.↗
DOE OSTI2021

POST-TEST EXAMINATIONS OF A LOCA SAMPLE FROM AN IRRADIATED HIGH-BURNUP PWR M5 FUEL ROD

A LOCA integral test with high burnup PWR M5 fuel rod was conducted in the Irradiated Fuel Examination Laboratory through the complete LOCA sequence: heating the LOCA sample to 300ºC and pressurizing the internal pressure to 8.27 MPa, heating at 5ºC/s from 300 to 1200ºC, holding in steam for 90s at 1200ºC, cooling at 3ºC/s to 800ºC, followed by water quench and rapid cooling to 100ºC. After LOCA testing, examinations, such as the fuel fragmentation analysis, burst and ballooning characterization, axial strain measurement, and microstructural examinations were performed. Metallographic examinations of an as-irradiated high burnup sample adjacent to the LOCA test sample revealed a bonding layer between the fuel and cladding. The posttest LOCA examinations indicates the corrosion layer formed during normal operations in the commercial reactor might provide a protection against the steam oxidation at high temperatures for test times performed in this work. The microstructure of the as-irradiated fuel is compared to the microstructure of the post-LOCA test fuel. Posttest LOCA examination on unirradiated post-tests Zr cladding samples was conducted, which served as baseline data for in cell testing with irradiated samples. The results obtained with irradiated PWR high burnup M5 fuel rad were compared to the LOCA test data obtained with irradiated BWR high burnup Zircaloy-2 fuel rod at Argonne national Laboratory.

Yan, Yong↗
NASA NTRS1975

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

Thoms, K. R.↗
NASA NTRS1977

Development of an experiment for determining the autoignition characteristics of aircraft-type fuels

An experimental test apparatus was developed to determine the autoignition characteristics of aircraft-type fuels in premixing prevaporizing passages at elevated temperatures and pressures. The experiment was designed to permit independent variation and evaluation of the experimental variables of pressure, temperature, flow rate, and fuel-air ratio. A comprehensive review of the autoignition literature is presented. Performance verification tests consisting of measurements of the ignition delay times for several lean fuel-air mixture ratios were conducted using Jet-A fuel at inlet air temperatures in the range 600 K to 900 K and pressures in the range 9 atm to 30 atm.

Spadaccini, L. J.↗
DOE OSTI2020

Fission Gas Monitoring System Efficiency Calibration Summary Report

The Fission Gas Monitoring System was designed to work in conjunction with the Fuel Accident Condition Simulator Furnace and is used to measure fission gas releases during post-irradiation fuel heating tests. A high-purity helium sweep gas is swept past the heated fuel sample in the furnace and routed to a fission gas monitoring system, which cryrogenically traps the krypton (Kr) and xenon (Xe) fission gases. As the radioactive species of the trapped gas decays, high-purity germanium detectors monitor the amount of fission gas in each trap. The primary fission product present in this test is 85Kr, which has a 10.76-year half-life. The presence of 85Kr indicates particle defects and iodine release. There is also the potential for re-irradiating the fuel compacts in the Neutron Radiography Reactor in the Hot Fuel Examination Facility basement. With re-irradiation, it is possible to have xenon fission gas present, specifically 133Xe, which has a 5.243-day half-life. Re-irradiation also produces 131I, which is important for design-basis accidents and regulatory licensing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AGR 3/4 Irradiation Test Final As Run Report, Revision 1

Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5 × 10 25 n/m 2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the temperature in the twelve capsules relatively flat in a range of temperatures suitable for the measurement of fission product diffusion in compact matrix and structural graphite materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

This is a conference presentation ppt file. The following abstract is for the entire conference paper - which will later be the subject of a PRS artifact The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗