Production upgrading and testing of fusel alcohols as high performance fuels from whole algae bioconversion.
Abstract not provided.
Monroe, Eric↗
SEARCH · Engineering Papers
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.
Production upgrading and testing of fusel alcohols as high performance fuels from whole algae bioconversion.
Abstract not provided.
Monroe, Eric↗
Performance of FeCrAl Accident Tolerant Fuel Cladding under Reactivity Insertion Accident Condition Testing
Explore the source record for details and available documents.
Garrison, Ben↗
Local Power Impact Experiment Design for a New Fuel Type for use in the Advanced Test Reactor
see INL/CON-23-73096 This LRS artifact is for the conference presentation associated with an abstract that was previously accepted through the LRS.
42 ENGINEERING↗
Field Testing of Safeguards Technologies in the Hot Fuel Examination Facility
Recent developments in nuclear fuel reprocessing techniques have yielded more efficient processes and fuel cycle options that strengthen the nuclear industry and production of clean energy. One such area of interest is pyroprocessing of used oxide fuel. However, with these advances in the back end of the nuclear fuel cycle, advances in safeguards instrumentation, measurements, and approaches are needed to ensure special nuclear material (SNM) is accounted for according to regulatory requirements. As a high-level overview of a nominal pyroprocessing approach, used oxide fuel from commercial light water reactors (LWR) is mechanically removed from the metallic cladding. Then the fuel is crushed and randomized representative samples are taken and sent to an analytical lab for analysis. The analytical results of the feed material are used for input accountancy into the rest of the process. The crushed oxide fuel is then moved to the oxide reduction (OR) furnace where it is reduced from an oxide to metallic form. The OR product is distilled to remove salt and then moved to an electrorefiner (ER), where it is immersed in a eutectic mixture of lithium chloride potassium chloride (LiCl-KCl) that typically ranges between 450-550 ?. Within the ER, the usable uranium is electrochemically transported through the molten salt from the anode to the cathode, and then subsequently removed as a relatively pure U product.. A simplified model of pyroprocessing techniques with added emphasis on the safeguards can be seen below in Fig. 1
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Evaluation for Effect of TRACE Spacer Grid Model on RBHT Reflood Tests
Usually, the spacer grid decreases the flow area and then enhances the convective heat transfer due to the flow acceleration and the turbulence increases. Their phenomena is shown in both egg-crate style spacer grids and mixing vane grids. Also, the mixing vane swirls the flow and enhances the convective heat transfer. In this study, the effect of spacer grid model of TRACE was performed on RBHT (Rod Bundle Heat Transfer) reflood tests. The RBHT tests were designed to conduct several separate-effects tests such as single phase steam cooling, low flow boiling and dispersed flow film boiling heat transfer, which is very important during a large break loss of coolant accident (LOCA). This facility has a heated length of 3.66 m (12.0 ft), which is typical of a pressurized water reactor (PWR) fuel assembly. The test section consists of a 7 x 7 rod bundle with 45 heated electrical rods, and seven mixing vane grids. The electrical rods have a diameter of 9.5 mm and are arranged with a 12.6 mm pitch with 45 heater rods and four unheated rods. Tests 1096, 1108, 1170, 1196, 1285 and 1383, which are in the TRACE assessment manual , were selected to evaluate the effect of the spacer grid model of TRACE. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Fuel-Flexible Gas Turbine Combustor Flametube Facility
Facility modifications have been completed to an existing combustor flametube facility to enable testing with gaseous hydrogen propellants at the NASA Glenn Research Center. The purpose of the facility is to test a variety of fuel nozzle and flameholder hardware configurations for use in aircraft combustors. Facility capabilities have been expanded to include testing with gaseous hydrogen, along with the existing hydrocarbon-based jet fuel. Modifications have also been made to the facility air supply to provide heated air up to 350 psig, 1100 F, and 3.0 lbm/s. The facility can accommodate a wide variety of flametube and fuel nozzle configurations. Emissions and performance data are obtained via a variety of gas sample probe configurations and emissions measurement equipment.
Little, James E.↗
Irradiation Testing of Sensors - ASI Program Review 2022
This is a powerpoint presentation for the annual Advanced Sensors and Instrumentation (ASI) program webinar. The following is a summary of the presentation. Advanced instrumentation enables testing of nuclear fuels and materials in support of the US advanced nuclear technology industry. A number of promising sensor technologies are in the pipeline. The early part of sensor development can be done outside of the reactor environment, but full technical readiness requires experience gained from in-core performance testing. Customers usually have only one shot to conduct their irradiation experiments. Because of the high costs test sponsors are frequently reluctant to incorporate unproven technologies in reactor experiments. Therefore, it is vital to demonstrate newly-developed sensors in operational conditions, prior to incorporating them into long-term high-value experiments. Successful completion of these activities will create new monitoring capabilities in high-power test reactors (such as ATR) as well as specialty reactors such as TREAT.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Technical Documents for Gateway for Accelerated Innovation in Nuclear (GAIN)
The Gateway for Accelerated Innovation in Nuclear (GAIN) provides the nuclear energy community with access to the technical, regulatory, and financial support necessary to move new or advanced nuclear reactor designs toward commercialization. GAIN provides the nuclear community with a single point of access to the broad range of capabilities (i.e., people, facilities, materials, and data) across the U.S. Department of Energy (DOE) complex and its National Lab capabilities. The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by DOE. The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following the demise of the breeder reactor program in the United States, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuable information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, reactor control, prototype testing, and site construction. This report provides documents related to three important aspects of FFTF design and operation: 1) irradiation behavior of structural alloys and absorber materials, 2) thermohydraulics of rod bundles (i.e., coolant mixing), and 3) natural circulation heat transfer in the areas of modeling and validation. These technical documents are believed to be of interest to the nuclear industry and in particular to designers of new liquid metal reactors.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Product Module Rig Test
The low emissions potential of a Rich-Quench-Lean (RQL) combustor for use in the High Speed Civil Transport (HSCT) application was evaluated as part of Work Breakdown Structure (WBS) 1.0.2.7 of the NASA Critical Propulsion Components (CPC) Program under Contract NAS3-27235. Combustion testing was conducted in cell 1E of the Jet Burner Test Stand at United Technologies Research Center. Specifically, a Rich-Quench-Lean combustor, utilizing reduced scale quench technology implemented in a quench vane concept in a product-like configuration (Product Module Rig), demonstrated the capability of achieving an emissions index of nitrogen oxides (NOx EI) of 8.5 gm/Kg fuel at the supersonic flight condition (relative to the program goal of 5 gm/Kg fuel). Developmental parametric testing of various quench vane configurations in the more fundamental flametube, Single Module Rig Configuration, demonstrated NOx EI as low as 5.2. All configurations in both the Product Module Rig configuration and the Single Module Rig configuration demonstrated exceptional efficiencies, greater than 99.95 percent, relative to the program goal of 99.9 percent efficiency at supersonic cruise conditions. Sensitivity of emissions to quench orifice design parameters were determined during the parametric quench vane test series in support of the design of the Product Module Rig configuration. For the rectangular quench orifices investigated, an aspect ratio (length/width) of approximately 2 was found to be near optimum. An optimum for orifice spacing was found to exist at approximately 0.167 inches, resulting in 24 orifices per side of a quench vane, for the 0.435 inch quench zone channel height investigated in the Single Module Rig. Smaller quench zone channel heights appeared to be beneficial in reducing emissions. Measurements were also obtained in the Single Module Rig configuration on the sensitivity of emissions to the critical combustor parameters of fuel/air ratio, pressure drop, and residence time. Minimal sensitivity was observed for all of these parameters.
Holdeman, James D.↗
Effect of some nitrogen compounds thermal stability of jet A
The effect of known concentrations of some nitrogen containing compounds on the thermal stability of a conventional fuel, namely, Jet A was investigated. The concentration range from 0.01 to 0.1 wt% nitrogen was examined. Solutions were made containing, individually, pyrrole, indole, quinoline, pyridine, and 4 ethylpyridine at 0.01, 0.03, 0.06, and 0.1 wt% nitrogen concentrations in Jet A. The measurements were all made by using a standard ASTM test for evaluating fuel thermal oxidation behavior, namely, ASTM D3241, 'thermal oxidation stability of turbine fuels (JFTOT procedure).' Measurements were made at two temperature settings, and 'breakpoint temperatures' were determined. The results show that the pyrrole and indole solutions have breakpoint temperatures substantially lower than those of the Jet A used.
Antoine, A. C.↗
2025 Intern Poster
The Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) houses the largest U.S. inert atmosphere hot cell for nuclear material research. Key features include the precision gamma scanning (PGS), Fuel Accident Condition Simulator (FACS), Neutron Radiography Reactor (NRAD), and the focus of this project, the Metallograph Loading Cell (MET Cell). The MET Cell performs tests on spent nuclear fuel, such as microhardness testing, microscopy, and neutron radiography. However, the MET Cell’s existing pressure and lighting control systems are outdated and inefficient, with inadequate documentation for system changes over time. This project aims to design a new automated control system for the MET Cell, ensuring longevity (minimum ten years), ease of troubleshooting/repair, and integration into the building monitoring system. The design process addressed challenges such as space restrictions, varied voltages within enclosures, sourcing new components, and security limitations. Compliance with NFPA 70, UL508A, MFC Physical Security, and INL Engineering standards was essential. The project involves repurposing an existing PLC to manage lighting and pressure control using digital and analog signals, simplifying wiring, and ensuring thorough documentation for future reference.
42 - ENGINEERING↗
Startup Test Plan and Predictions for Highly Enriched Uranium to Low-Enriched Uranium Fuel Conversion at the University of Missouri Research Reactor
Nonpower reactors licensed by the U.S. Nuclear Regulatory Commission require a startup test plan as part of any facility modification to verify operability prior to resumption of operations. In order to support conversion of the University of Missouri Research Reactor from the use of highly enriched uranium to low-enriched uranium (LEU) fuel, a startup test plan has been devised to measure certain reactor physics parameters for the initial all-fresh LEU core licensing documentation that will be submitted. These parameters include the approach to critical, primary coolant void coefficient of reactivity, flux trap void coefficient of reactivity, determination of flux trap sample reactivity worth, radial and axial thermal neutron flux mapping, control blade worth calibration, primary and pool coolant temperature coefficient of reactivity, and flux mapping of experimental positions. Here, predictions for these parameters made using the Monte Carlo N-Particle Version 5 (MCNP5) radiation transport code are reported. These predictions will support the startup tests by providing a baseline set of expectations and additional insight into the performance of the LEU core.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FFTF Acceptance and Startup Testing for GAIN
The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following the demise of the breeder reactor program in the United States, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. FFTF operations ceased in April 1992 after a determination by DOE that no combination of proposed missions was financially feasible over a ten-year period. The reactor is currently deactivated and in a long-term surveillance and maintenance (S&M) mode. This report provides information on the extensive and rigorous process that was used to conduct turnover from construction followed by acceptance and startup testing of the FFTF. This paper is in support of the Gateway for Accelerated Innovation in Nuclear (GAIN), which provides the nuclear energy community with access to the technical, regulatory, and financial support necessary to move new or advanced nuclear reactor designs toward commercialization while ensuring the continued safe, reliable, and economic operation of the existing nuclear fleet. The information obtained from the design, startup, and operation of the FFTF provides valuable insight for follow-on reactor projects, such as the Versatile Test Reactor (VTR), in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, reactor startup and operations, and reactor deactivation and decommissioning (D&D). The focus of this report is on the process used to startup the FFTF and to ensure that operations could be conducted efficiently and safely. A reference section is provided of documents detailing the successful turnover and testing process implemented for startup of the reactor and its supporting systems. The documents listed can be retrieved upon request and are believed useful for future reactor startup endeavors.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Shaker Table Test Plan
Currently, spent nuclear fuel (SNF) is stored in onsite independent spent fuel storage facilities (ISFSIs), which is a dry storage facility, at 55 nuclear power plant sites. The majority of SNF in dry storage is in welded metal canisters (2,917 canisters at the end of 2019). The canisters are loaded for storage in storage overpacks (vertical casks or horizontal storage modules) and placed on outdoor concrete pads. Because the SNF will be stored at ISFSIs for an extended period of time, there is growing concern with regards to the behavior of the SNF within these dry storage systems during earthquakes. To address these concerns, the SFWST program is considering conducting an earthquake shaker table test. The goal of this test is to determine the strains and accelerations on fuel assembly hardware and cladding during earthquakes of different magnitudes to better quantify the potential damage an earthquake could inflict on spent nuclear fuel rods. The seismic integrity of the storage system has been addressed in the past by the US Nuclear Regulatory Commission and is not the focus of this potential test. Instead the DOE would benefit from knowing the condition of the fuel cladding from storage, transportation, to disposal so that it can ascertain repository performance for the fuel and packaging in its final state. A seismic event is part of the possible loading events that the fuel could experience in its lifetime. This report proposes several earthquake shaker table tests with different degrees of complexity. Alternative 1 was defined in the FY20 work scope. Alternatives 2 and 3 were recently developed to take advantage of the NUHOMS 32PTH dry storage canister that may be available in FY21 for this test at a minimum cost to the project. The selection of the alternative(s) will depend on the available budget and the SFWST program priorities for the near future.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Validation of the Fast and Thermal Neutron Flux Profiles in the Advanced Test Reactor for the AGR-3/4 Experiment
The AGR-3/4 experiment was designed to irradiate TRISO particle fuel in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. Temperature control of the test capsules containing the TRISO particles was important since the goal was to measure diffusion rates of fission product elements through graphite as a function of temperature. During the experiment, differences were observed between temperatures measured by the thermocouples and temperatures calculated with the thermal analysis model. To rule out the possibility that the used ATR driver fuel elements were the root cause of these temperature differences, an experiment was devised to measure the axial fast and thermal neutron flux profiles in the northeast flux trap containing the AGR-3/4 experiment. An important input parameter to the thermal analysis model was the calculated compact heat rates from the MCNP physics model. To verify these calculated heat rates were consistent with the ATR driver fuel element loading and operating lobe power, axial flux wires were inserted into ATR northeast flux trap. The measured fast and thermal neutron axial flux profiles derived from the flux wire activation showed excellent agreement with the calculated flux profiles. This agreement dispelled any concerns that the used ATR fuel elements in the northeast lobe might have introduced unusual axial flux variation. In addition, the agreement provided a validation basis for the physics calculation, specifically, (1) MCNP physics model, (2) Monte Carlo depletion methodology, (3) method used to load the ATR driver fuel, and (4) ATR measured lobe power data used in the calculation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Dry in-pile fracture test (DRIFT) for separate-effects validation of ceramic fuel fracture models
Fracture is an important component of nuclear fuel behavior, and significant efforts have been invested into developing fuel performance models that are capable of accurately representing fracture. Usable data on the process of fracture propagation in nuclear fuel under realistic service conditions are very limited. To address this need, a series of separate-effects experiments were developed and performed at Idaho National Laboratory's Transient Reactor Test (TREAT) facility. These experiments employ a heat sink to radially remove heat from the fuel in a manner that approximates the effect of coolant in an operating light-water reactor (LWR). The test holder for these experiments is known as the Dry In-pile Fracture Test (DRIFT). A series of experiments employing DRIFT and TREAT were performed to provide data on the extent and nature of fracture in fresh fuel at various points during a ramp to full power. Novel aspects of these experiments include the way they employ a heat sink to replicate steady-state LWR conditions, as well as the use of fiber optic sensors for in-reactor thermal instrumentation. Details on the development of this experiment, experimental conditions, and resulting data (including in situ thermal measurements and post-irradiation imaging of fracture) are provided in this work. LWR-equivalent powers ranging from 10 to 25 kW/m were tested using this apparatus. Cracking was visible at all power levels, with increasing cracking extent as the power level increased, although there was little difference in the cracking between the two highest-power tests, which had LWR-equivalent powers of 20 and 25 kW/m.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Small global effect on terrestrial net primary production due to increased fossil fuel aerosol emissions from East Asia since the turn of the century
Abstract The global terrestrial carbon sink has increased since the start of this century at a time of growing carbon emissions from fossil fuel burning. Here we test the hypothesis that increases in atmospheric aerosols from fossil fuel burning enhanced the diffuse light fraction and the efficiency of plant carbon uptake. Using a combination of models, we estimate that at global scale changes in light regimes from fossil fuel aerosol emissions had only a small negative effect on the increase in terrestrial net primary production over the period 1998–2010. Hereby, the substantial increases in fossil fuel aerosol emissions and plant carbon uptake over East Asia were effectively canceled by opposing trends across Europe and North America. This suggests that if the recent increase in the land carbon sink would be causally linked to fossil fuel emissions, it is unlikely via the effect of aerosols but due to other factors such as nitrogen deposition or nitrogen‐carbon interactions.
54 ENVIRONMENTAL SCIENCES↗