Search NASA⌕ Search

SEARCH · Search NASA

Results for “Fuel testing”

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.

At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two primary goals. First, the test was intended to assess the retention of fission products in loose kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material comprising the compact. Second, this test served as an evaluation of the FACS fission product condensation plate collection efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Enhancing TREAT’s Pulsing Capabilities Using TREAT Upgrade Fuel

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is an air-cooled, thermal, heterogeneous facility used to test reactor materials in simulated accident conditions by inducing fission heating with intense neutron pulses. TREAT operated from 1959 until 1994 with the primary goal of testing fast reactor fuels and was brought back online in 2017 to re-establish DOE's nuclear fuels transient testing capabilities. Beginning in the late 1970s, Idaho National Laboratory worked in conjunction with other organizations to increase TREAT's capability for in-pile testing. While these upgrades had not been implemented by the time the reactor was put into standby mode, new assemblies and graphite-urania fuel blocks with increased uranium concentrations had already been designed and fabricated. This project, as part of Idaho National Laboratory's Department of Reactor Physics, modeled the TREAT Upgrade (TU) fuel assemblies in MCNP and implemented them in an existing TREAT model. Calculations were performed using INL's high performance computer to find a critical combination of TU and standard fuel and this geometry's excess reactivity. Future work will include calculating a power coupling factor in the experiment, determining power peaking factors throughout the core, and assessing how much reactivity would be added by replacing the Inconel-625 cladding with silicon carbide.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1974

Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K

Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.

Diianni, D. C.↗
DOE OSTI2018

Membrane Accelerated Stress Test Development for Polymer Electrolyte Fuel Cell Durability Validated Using Field and Drive Cycle Testing

A combined chemical/mechanical accelerated stress test (AST) was developed for proton exchange membrane (PEM) fuel cells based on relative humidity cycling (RHC) between dry and saturated gases at open circuit voltage (OCV). Membrane degradation and failure were investigated using scanning electron microscopy and small- and wide-angle X-ray scattering. Changes to membrane thickness, hydrophilic domain spacing, and crystallinity were observed to be most similar between field-operated cells and OCV RHC ASTs, where local thinning and divot-type defects are the primary failure modes. While RHC in air also reproduces these failure modes, it is not aggressive enough to differentiate between different membrane types in >1,333 hours (55 days) of testing. Conversely, steady-state OCV tests result in significant ionomer morphology changes and global thinning, which do not replicate field degradation and failure modes. It is inferred that during the OCV RHC AST, the decay of the membrane's mechanical properties is accelerated such that materials can be evaluated in hundreds, instead of thousands, of hours, while replicating the degradation and failure modes of field operation; associated AST protocols are recommended as OCV RHC at 90°C for 500 hours with wet/dry cycle durations of 30s/45s and 2m/2m for automotive and bus operation, respectively.

25 ENERGY STORAGE↗
DOE OSTI2017

Durability Testing of Biomass Based Oxygenated Fuel Components in a Compression Ignition Engine

Blending cellulosic biofuels with traditional petroleum-derived fuels results in transportation fuels with reduced carbon footprints. Many cellulosic fuels rely on processing methods that produce mixtures of oxygenates which must be upgraded before blending with traditional fuels. Complete oxygenate removal is energy-intensive and it is likely that such biofuel blends will necessarily contain some oxygen content to be economically viable. Previous work by our group indicated that diesel fuel blends with low levels (<4%-vol) of oxygenates resulted in minimal negative effects on short-term engine performance and emissions. However, little is known about the long-term effects of these compounds on engine durability issues such as the impact on fuel injection, in-cylinder carbon buildup, and engine oil degradation. In this study, four of the oxygenated components previously tested were blended at 4%-vol in diesel fuel and tested with a durability protocol devised for this work consisting of 200 hrs of testing in a stationary, single-cylinder, Yanmar diesel engine operating at constant load. Oil samples, injector spray patterns, and carbon buildup from the injector and cylinder surfaces were analyzed. It was found that, at the levels tested, these fuels had minimal impact on the overall engine operation, which is consistent with our previous findings.

09 BIOMASS FUELS↗
DOE OSTI2016

Modelling of LOCA Tests with the BISON Fuel Performance Code

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

BISON code↗
NASA NTRS2010

Thermal Stability Testing of a Fischer-Tropsch Fuel and Various Blends with Jet A

Fischer-Tropsch (F-T) jet fuel composition differs from petroleum-based, conventional commercial jet fuel because of differences in feedstock and production methodology. Fischer-Tropsch fuel typically has a lower aromatic and sulfur content and consists primarily of iso and normal parafins. The ASTM D3241 specification for Jet Fuel Thermal Oxidation Test (JFTOT) break point testing method was used to test the breakpoint of a baseline conventional Jet A, a commercial grade F-T jet fuel, and various blends of this F-T fuel in Jet A. The testing completed in this report was supported by the NASA Fundamental Aeronautics Subsonics Fixed Wing Project.

Klettlinger, Jennifer Suder↗
DOE OSTI2019

Preliminary Evaluation of Fuel Management Strategies for the Versatile Test Reactor (VTR)

Different fuel management approaches have been investigated for the Versatile Test Reactor (VTR). The main design variables were the number of batches, the number of (orifice) zones, and the loading pattern within the core. The optimization criteria are maximizing the fast flux at the centermost position, minimizing the peak assembly power, and reducing fuel consumption. Design constraints were mainly driven by material limits such as peak clad temperature, and maximum fuel burnup. The exploratory study determined that active fuel shuffling only brought very limited benefits. A three-orifice region was found to provide enough flexibility to ensure thermal criteria were met. Lastly, a flexible batch loading termed ‘5-6-8’ was identified as a promising management strategy from a resource utilization standpoint.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1982

Low NO sub x heavy fuel combustor concept program. Phase 1A: Combustion technology generation coal gas fuels

Combustion tests of two scaled burners using actual coal gas from a 25 ton/day fluidized bed coal gasifier are described. The two combustor configurations studied were a ceramic lined, staged rich/lean burner and an integral, all metal multiannual swirl burner (MASB). The tests were conducted over a range of temperature and pressures representative of current industrial combustion turbine inlet conditions. Tests on the rich lean burner were conducted at three levels of product gas heating values: 104, 197 and 254 btu/scf. Corresponding levels of NOx emissions were 5, 20 and 70 ppmv. Nitrogen was added to the fuel in the form of ammonia, and conversion efficiencies of fuel nitrogen to NOx were on the order of 4 percent to 12 percent, which is somewhat lower than the 14 percent to 18 percent conversion efficiency when src-2 liquid fuel was used. The MASB was tested only on medium btu gas (220 to 270 btu/scf), and produced approximately 80 ppmv NOx at rated engine conditions. Both burners operated similarly on actual coal gas and erbs fuel, and all heating values tested can be successfully burned in current machines.

Sherlock, T. P.↗
DOE OSTI2024

Specifications of Cladding Diameter Measurements Conducted at AGHCF

Fuel element diameter data was collected on-site at the Alpha-Gamma Hot Cell Facility (AGHCF) before and after out-of-pile furnace transient tests of fuel elements. Available data records have been collected and preserved in the Out-of-Pile Transient Database (OPTD). This data is used to determine the transient-induced changes in fuel element diameter, or cladding strain, for the Whole Pin Furnace (WPF) tests. Fuel element diameter was measured by contact profilometry along the length of the fuel pin at specified rotational orientations and/or by a manually operated micrometer at several discrete axial locations along the pin. The Alpha-Gamma Hot Cell Facility Operations Manual details the way the facility operated, organizational and oversight responsibilities, and procedures for examination of samples. The working version of the manual at the time of the whole pin furnace tests and the test pin examinations is Doc. No. IPS-2-00-00, dated June 1989. This specification was developed using the operations manual and recovered measurement records in consultation with subject matter experts (SMEs). The measurement methods, format of the available post-test examination (PTE) data, and recommended methods for use and interpretation of that data are summarized. Section 2 describes the available diameter measurement data for the WPF tests with guidance for interpretation and usage, and Section 3 describes the instrument measurement procedures and calibration methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multiphysics Simulation of the NASA SIRIUS-CAL Fuel Experiment in the Transient Test Reactor Using Griffin

After approximately 50 years, NASA is restarting efforts to develop nuclear thermal propulsion (NTP) for interplanetary missions. Building upon nuclear engine tests performed from the late 1950s to the early 1970s, the present research and testing focuses on advanced materials and fabrication methods. A number of transient tests have been performed to evaluate materials performance under high-temperature, high-flux conditions, with several more experiments in the pipeline for future testing. The measured data obtained from those tests are being used to validate the Griffin reactor multiphysics code for this particular type of application. Griffin was developed at Idaho National Laboratory (INL) using the MOOSE framework. This article describes the simulation results of the SIRIUS-CAL calibration experiment in the Transient Reactor Test Facility (TREAT). SIRIUS-CAL was the first transient test conducted on NASA fuels, and although the test was performed with a relatively low core peak power, the test specimen survived a temperature exceeding 900 K. Griffin simulations of the experiment successfully matched the reactor’s power transient after calibrating the initial control rod position to match the initial reactor period. The thermal-hydraulics model largely matches the time-dependent response of a thermocouple located within the experiment specimen to within the uncertainty estimate. However, the uncertainty range is significant and must be reduced in the future.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2018

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗