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 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
NASA NTRS1995

Metallized Gelled Propellants: Oxygen/RP-1/aluminum Rocket Combustion Experiments

A series of combustion experiments were conducted to measure the specific impulse, Cstar-, and specific-impulse efficiencies of a rocket engine using metallized gelled liquid propellants. These experiments used a small 20- to 40-1bf (89- to 178-N) thrust, modular engine consisting of an injector, igniter, chamber and nozzle. The fuels used were traditional liquid RP-1 and gelled RP-1 with 0-, 5-, and 55-wt% loadings of aluminum and gaseous oxygen was the oxidizer. Ten different injectors were used during the testing: 6 for the baseline 02/RP-1 tests and 4 for the gelled fuel tests which covered a wide range of mixture ratios. At the peak of the Isp versus oxidizer-to-fuel ratio (O/F) data, a range of 93 to 99% Cstar efficiency was reached with ungelled 02/RP-1. A Cstar efficiency range of 75 to 99% was obtained with gelled RP-l (0-wt% RP-1/Al) while the metallized 5-wt% RP-1/Al delivered a Cstar efficiency of 94 to 99% at the peak Isp in the O/F range tested. An 88 to 99% Cstar efficiency was obtained at the peak Isp of the gelled RP1/Al with 55-wt% Al. Specific impulse efficiencies for the 55-wt% RP-1/Al of 67%-83% were obtained at a 2.4:1 expansion ratio. Injector erosion was evident with the 55-wt% testing, while there was little or no erosion seen with the gelled RP-1 with 0- and 5-wt% Al. A protective layer of gelled fuel formed in the firings that minimized the damage to the rocket injector face. This effect may provide a useful technique for engine cooling. These experiments represent a first step in characterizing the performance of and operational issues with gelled RP-1 fuels.

Palaszewski, Bryan↗
DOE OSTI2015

Synthesis and Characterization of Mixed-Conducting Corrosion Resistant Oxide Supports

An extensive search and evaluation of electrochemically stable catalyst supports (including metal oxides like RuO2-SiO2, RuO2-TiO2, and ITO was perfomed during the 4 years of the project. The suports were also catalyzed by deposition of Pt and tested for its performance and electrochemical stability in RDE and fuel cell experiments. For testing the electrochemical stability and fuel cell performance of the catalysts and supports, we have employed the protocols in use at the Department of Energy and Nissan Technological Center North America (NTCNA). The use of such procedures allows a precise and reproducible estimation of the performance and stability of the materials and permits comparisons among laboratories and DOE funded projects. RuO2-SiO2 catalyst supports showed no loss in surface area during start-stop stability tests that were performed by cycling the electrode potential between 0 V to 1.8 V for 1000 cycles. Catalyzed support (40% Pt/RuO2-SiO2; 1:1 mole ratio) were tested in a PEFC, resulting in a current density of 750 mA/cm2 at 0.6 Volts, and a maximum power density of 570 mW/cm2. Measurements were conducted at 80 ºC with 75% relative humidity of the inlet gases (H2/O2); Pt loadings were 0.4 mg/cm2 at the cathode and 0.2 mg/cm2 at the anode. Start-stop stability tests for support and catalyzed support performed in RDE and PEFC set-ups have confirmed RuO2-TiO2 support stability. The beginning of life performance was exactly equal to end of life performance (in an MEA that has been subjected to severe start-stop cycling for 10,000 start/stop cycles between 1 V to 1.5 V). This result was in sharp contrast to baseline Pt/C catalyst that showed significant performance deterioration after accelerated stability tests. The Pt/TRO showed minimal loss in performance upon exposure to start-stop cycles. The loss in cell voltage at 1 A/cm2 at 100% RH was almost 700 mV for Pt/C whereas it was only ca. 15 mV for Pt/TRO. 40% RH data (of inlet gases) revealed a similar trend in terms of stability – exceptional stability for Pt/TRO as opposed to very poor stability for Pt/HSAC. These observations were attributed to the much higher stability of the TRO support compared to Carbon. The carbon dioxide concentration in the cathode exit stream during the accelerated degradation test with Pt/TRO (start-stop protocol) was extremely low (between 3 to 10 ppm of CO2). In contrast, the CO2 emission levels from a conventional Pt/C catalyst were found to be approx. 200 ppm. This observation was a clear indicator that the main source of carbon being oxidized to carbon dioxide in an MEA was the carbon catalyst support, and not the gas diffusion layer or the graphite flow fields. Indium tin oxide (ITO) was also evaluated as a catalyst support for PEFCs. Pt/ITO was very stable under start-up/shutdown accelerated degradation protocol (RDE tests in perchloric acid). The ECSA change was less than 4% over 10,000 cycles. The load cycling accelerated protocol (from 0.6 to 0.95 V vs. RHE) resulted in a loss of approximately 34% of the initial ECSA after 10,000 cycles. However, fuel cell testing resulted in a very low performing catalyst. XPS spectroscopy was employed to investigate the changes in the catalysts occuring during fuel cell operation. It was observed a shift of In 3d5/2 and In 3d3/2 peaks towards higher binding energies. This can be explained by the formation of hydroxides or oxy-hydroxides in the surface of the catalyst. O1s spectrum for Pt/ITO catalyst after being operated in the fuel cell, also confirmed the formation of significant amounts of surface hydroxides (12 to 16%). The presence of surface hydroxides in the catalyst increased the electrode resistivity affecting fuel cell performance. NTCNA performed a detailed analysis of transport phenomena (reactants and products to/from the Pt active sites) in both commercial catalyst and Pt/RTO (in order to have a better understanding at the basic level). The proton resistance (Rionomer) in Pt/C and Pt/RTO cathode catalyst layers were 150 and 12 mΩ-cm2, respectively. Pt/RTO catalyst layer has about an order or magnitude lower proton transfer resistance than Pt/C catalyst layer. Since the ionomer/support ratio that was used in formulating the ink for both catalysts was the same (0.9), it is expected that the volumetric coverage of ionomer of both catalysts will be significantly different due to the disparity in the surface areas (Pt/C had ~ 800 m2/g, while Pt/RTO had ~ 50 m2/g). The differences in the ionomer volumetric coverage and the ionomer film thickness may explain the significantly higher proton conductivity in the Pt/RTO catalyst layer when compared to Pt/HSAC. It is therefore very important to optimize the ionomer loadings when synthesizing new catalyst supports (and never rely on values for carbon-based commercial catalysts). Finally, NTCNA has elaborated a cost model for non-carbon support materials considering their durability benefits. Material costs for production of Pt/ RuO2-TiO2 electrodes were compared to Pt/C. RuO2-TiO2 support was more expensive than carbon but the total material cost was still dominated by platinum cost. Though ruthenium is considered a precious metal, its cost is far less than platinum. It should also be noted that ruthenium only makes up 38% of the mass of the support, while the rest is inexpensive TiO2. After considering the durability advantages of Pt/RTO, cost model showed that even with almost double the Pt loading (0.35 vs 0.18 mgPt/cm2), Pt/RTO ($22.7/kWnet) is only slightly more expensive than Pt/C ($21.9/kWnet).

08 HYDROGEN↗
DOE OSTI2018

Emissions of particulate matter, carbon monoxide and nitrogen oxides from the residential burning of waste paper briquettes and other fuels

Highlights: • The first attempt to report emission of key air pollutants from waste paper briquettes. • Burning of paper briquettes emitted relatively high CO and NO{sub x} concentration. • Burning of paper briquettes emitted PM with large average particle size. • Waste paper could be recycled as briquettes for domestic use like other common fuels. Using waste paper as fuel for domestic heating is a beneficial recycling option for small island developing states where there are lacks of resources for energy and waste treatment. However, there are concerns about the impact of air pollutants emitted from the burning of the self-made paper briquettes as household air pollution is recognised as the greatest environmental risk for human. In this study, combustion tests were carried out for paper briquettes made in one Pacific island and three commercial fuels in Australia including wood briquettes, kindling firewood and coal briquettes in order to: 1) characterise the emissions of three criteria air pollutants including particulate matters, CO and NO{sub x} including their emission factors (EF) from the tested fuels; and 2) compare the EFs among the tested fuels and with others reported in the literature. The results showed that waste paper briquettes burned quickly and generated high temperature but the heat value is relatively low. Paper briquettes and coal briquettes produced higher CO concentration than the others while paper briquettes generated the highest NO{sub x} level. Only PM{sub 2.5} concentration emitted from paper briquettes was similar to kindling firewood and lower than wood briquettes. Burning of paper briquettes and wood briquettes produced particulate matter with large average count median diameter (72 and 68 nm) than coal briquette and kindling firewood (45 and 51 nm). The EFs for CO, NO{sub x} and PM{sub 2.5} of paper briquettes were within the range of EFs reported in this study as well as in the literature. Overall, the results suggested that using paper briquettes as fuel for domestic heating will not likely to generate higher level of three major air pollutants compared to other traditional fuels.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

Equibiaxial flexural strength determination of UO 2 using a ball-on-ring test

To increase nuclear fuel performance and reliability, their mechanical properties require an accurate and statistically relevant assessment. Biaxial flexural strength tests provide an alternative to bend bar techniques for assessing mechanical behavior; namely, the transverse rupture strength (TRS) of ceramic samples. Biaxial test samples require simple geometries and minimal surface preparation, reducing fabrication costs and handling hazards. This study investigated the TRS of polycrystalline UO 2 fuel forms at room temperature using a ball-on-ring test fixture. Here, pellets were fabricated from UO 2 powder using conventional powder processing and sintering techniques. The TRS and Weibull parameters were obtained through Weibull statistics on over 60 UO 2 samples tested under equibiaxial flexure. The larger sample size in this study enabled a more robust Weibull statistical analysis than alternative test methods, which may not capture the stochastic failure of typical ceramics. Furthermore, two different loading ball diameters were employed to assess the impact of contact damage on fracture strength. While Hertzian contact damage was observed with the smaller loading ball, the fracture strength remained unaffected. A fracture analysis of the tested UO 2 samples indicated a mixture of intergranular and transgranular fracture that transitions to transgranular fracture with increasing distance from the fracture origin. The characteristic strength of the combined data sets was determined to be 148 MPa, and the Weibull modulus was determined to be 9.1. The TRS values and Weibull parameters were close to values found in the literature for alternative testing techniques using samples with similar microstructure and density. The findings in this study validate the ball-on-ring method used to obtain the TRS of UO 2 with a sample geometry more representative of nuclear fuels. Additionally, experimental TRS results from this study can be implemented in modeling codes to predict fuel performance, which is critical to fuel burnup extension and advanced nuclear fuel technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix

The Transformational Challenge Reactor (TCR) fuel form was designed to contain large, densely packed uranium carbonitride (UCN) tristructural-isotropic (TRISO) fuel particles within a 3D printed SiC matrix, increasing the uranium density compared to conventional TRISO fuel forms and offering full geometric freedom for core design. Here, this work summarizes initial low-burnup, high-power irradiation testing of TCR fuel materials, including loose UCN TRISO particles and integral fuel compacts with ~55% TRISO particles by volume, to evaluate fission gas retention. Fission gasses were fully retained in all loose particle tests and in integral compacts irradiated at low (<250 °C) surface temperatures. Initial testing at higher (~700–750 °C) fuel surface temperatures showed fission gas release (FGR) and complete fracture of three compacts, but no FGR was observed in later high temperature tests (~300–750 °C) of both fueled compacts and loose TRISO particles. Calculated thermal stresses in the failed compacts were far less than the measured strength of the SiC matrix and the stresses in some failed compacts were less than those in compacts that did not show FGR. Thermal stress-induced matrix cracks also would not cause complete fracture because the tensile stresses transition to compression in the higher temperature regions. Therefore, fuel failure was likely not caused by thermal stresses and may have been related to leakage currents from the electrical heaters and erratic fuel surface temperatures that were only observed in the test for which failure was observed. In any case, the matrix cracks propagated through the coatings of TRISO particles located in the high-density matrix regions on the peripheries of the compacts, resulting in measurable fission gas release. The discussion focuses on the importance of understanding matrix density distributions and the particle-matrix interface properties to prevent matrix cracks from causing TRISO particle failures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Uranium and cesium diffusion in fuel cladding of electrogenerating channel

The results of reactor tests of a carbonitride fuel in a single-crystal cladding from a molybdenum-based alloy can be used in substantiating the operational reliability of fuels in developing a project of a megawatt space nuclear power plant. The results of experimental studies of uranium and cesium penetration into the single-crystal cladding of fuel elements with a carbonitride fuel are interpreted. Those fuel elements passed nuclear power tests in the Ya-82 pilot plant for 8300 h at a temperature of about 1500°C. It is shown that the diffusion coefficients for uranium diffusion into the cladding are virtually coincident with the diffusion coefficients measured earlier for uranium diffusion into polycrystalline molybdenum. It is found that the penetration of uranium into the cladding is likely to occur only in the case of a direct contact between the cladding and fuel. The experimentally observed nonmonotonic uranium-concentration profiles are explained in terms of predominant uranium diffusion along grain boundaries. It is shown that a substantially nonmonotonic behavior observed in our experiment for the uranium-concentration profile may be explained by the presence of a polycrystalline structure of the cladding in the surface region from its inner side. The diffusion coefficient is estimated for the grain-boundary diffusion of uranium. The diffusion coefficients for cesium are estimated on the basis of experimental data obtained in the present study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard-fired alumina insulation and a molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple, based on molybdenum/niobium alloys and developed at Idaho National Laboratory, was also tested. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The Impact of Furnace-Testing on the Microstructure of an Irradiated U-Mo Fuel Plate

The Material Management and Minimization Program (formerly known as the Reduced Enrichment for Research and Test Reactor Program) is responsible for converting research reactors that use highly enriched uranium (HEU) fuels to ones that use low-enriched uranium (LEU) fuels. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. In addition to the reactor tests, there is interest in generating information about the transient behavior of irradiated fuel plates. For example, it is of interest to determine the temperature at which a blister will form on the surface of a fuel plate, due to fission gas release from the fuel meat, for a particular fuel plate design. Since in-pile testing to determine the 'blister threshold' (the temperature where a blister forms on the surface of a fuel plate) in plate-type fuel is very difficult and expensive and has historically been deemed impractical for fuel qualification testing, an out-of-pile 'blister test' has been developed and employed. This post-irradiation blister anneal test was developed as a conservative simulation of in-pile behavior and has been routinely used for many research reactor fuel designs currently in use. In this test, the irradiated fuel plate is inserted into a furnace at a given temperature, soaked for a period of time, and removed for inspection. If no visible blisters appear, the furnace temperature is increased, and the plate is soaked and inspected again. This process continues until blisters are observed. Recently, a section from an irradiated U-7Mo dispersion fuel plate with Al-4043 matrix, which has a nominal composition of 4.81Si-0.20Fe-0.14Ti-0.16Cu-0.01Cr 0.01Mn-bal Al, was tested in a furnace located in the Hot Fuel Examination Facility, located at the Idaho National Laboratory, up to 500 deg. C, at which point blisters were observed on the surface of the fuel plate section. A sample near the blistered region of the fuel plate section was then characterized using scanning electron (SEM) and transmission electron microscopy (TEM). This paper discusses the results of this characterization and how they compare to those generated for a similar sample characterized after fabrication but before irradiation and for a sample characterized after irradiation but before furnace testing. One goal of this work is to identify the changes in the microstructure during blister testing that contribute to the eventual release of fission gas that results in the development of blisters. SEM analysis of the furnace-tested sample showed that in some locations of the U-7Mo particles, Al-4043 matrix constituents (primarily Si) had penetrated the microstructure along grain boundaries. This provided potential pathways for increased fission gas migration in the fuel meat microstructure during the furnace heating process. Some attack of these boundaries was also observed in the as-fabricated U-7Mo/Al-4043 sample, but there appeared to be additional attack during the furnace annealing. Besides the changes observed in grain boundary features, the TEM analysis showed that the fission gas bubbles that are typically present in the intra-granular regions of the as-irradiated U-7Mo microstructure as a fission gas bubble superlattice had changed in size, morphology, and distribution. The result of this may be an increase in the mobility of the fission gas in intra-granular regions of the fuel. The complete implications of the changes in fission gas morphology and distribution in the fuel meat microstructure and their impact on blistering continues to be investigated. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Lower-Energy Energy Storage System (LEESS) Component Evaluation

Alternate hybrid electric vehicle (HEV) energy storage systems (ESS) such as lithium-ion capacitors (LICs) and electrochemical double-layer capacitor (EDLC) modules have the potential for improved life, superior cold temperature performance, and lower long-term cost projections relative to traditional battery storage systems. If such lower-energy ESS (LEESS) devices can also be shown to maintain high HEV fuel savings, future HEVs designed with these devices could have an increased value proposition relative to conventional vehicles. NREL's vehicle test platform is helping validate the in-vehicle performance capability of alternative LEESS devices and identify unforeseen issues. NREL created the Ford Fusion Hybrid test platform for in-vehicle evaluation of such alternative LEESS devices, bench testing of the initial LIC pack, integration and testing of the LIC pack in the test vehicle, and bench testing and installation of an EDLC module pack. EDLC pack testing will continue in FY15. The in-vehicle LIC testing results suggest technical viability of LEESS devices to support HEV operation. Several LIC configurations tested demonstrated equivalent fuel economy and acceleration performance as the production nickel-metal-hydride ESS configuration across all tests conducted. The lowest energy LIC scenario demonstrated equivalent performance over several tests, although slightly higher fuel consumption on the US06 cycle and slightly slower acceleration performance. More extensive vehicle-level calibration may be able to reduce or eliminate these performance differences. The overall results indicate that as long as critical attributes such as engine start under worst case conditions can be retained, considerable ESS downsizing may minimally impact HEV fuel savings.

25 ENERGY STORAGE↗
NASA NTRS1982

Fuel economy and exhaust emissions characteristics of diesel vehicles: Test results of a prototype fiat 131TC 2.4 liter automobile

The results obtained from fuel economy and emission tests conducted on a prototype Fiat 131 turbocharged diesel vehicle are presented. The vehicle was tested on a chassis dynamometer over selected drive cycles and steady-state conditions. Two fuels were used, a United States number 2 diesel and a European diesel fuel. Particulate emission rates were calculated from dilution tunnel measurements and large volume particulate samples were collected for biological and chemical analysis. It was determined that turbocharging accompanied by complementary modifications results in small but substantial improvements in regulated emissions, fuel economy, and performance. Notably, particulate levels were reduced by 30 percent.

Quayle, S. S.↗
DOE OSTI2019

Preliminary Evaluation of Loading DOE Standardized Canisters in the CPP-603 Irradiated Fuel Storage Facility

This evaluation looks at the equipment and operations necessary to load United States Department of Energy (DOE) -owned Spent Nuclear Fuel (SNF) into DOE SNF Standardized Canisters in the CPP-603 Irradiated Fuel Storage Facility in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The ATR fuel ready for loading would come from fuel storage canisters in the CPP-603 facility, while the Peach Bottom fuel ready for loading could come from either fuel storage canisters in the CPP-603 facility and or containers in the CPP-749 Underground Fuel Storage Facility – Generation I Vaults.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Refabriation of HERA-HBU-2, LOC-HBU-1, and LOC-HBU-2

The development of advanced fuels for Light Water Reactors (LWRs) is essential for improved fuel performance and safety, including performance margins, uprates, and higher burnup. Because available commercial LWR fuels often cannot be used in research reactors, the refabrication of preirradiated fuel is necessary to enable follow-on testing of these fuels for characterizing fuel behavior during steady and transient irradiations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1980

Fuel character effects on the J79 and F101 engine combustion systems

The effects of select fuel property variations on two major engine classifications are summarized. Thirteen refined and blended fuels were used which exhibited significant variations in hydrogen content, aromatic type, initial boiling point, final boiling point, and viscosity. Trends were very similar but the degree of fuel sensitivity was not constant. For both systems the dominant fuel property during high pressure operation was found to be fuel hydrogen content. For operation at low pressure test points the fuel volatility and viscosity became the dominant fuel properties for both systems. Aromatic type and final boiling point did not significantly affect combustion data. Correlations of other fuel properties with these and other performance parameters are presented.

Jackson, T. A.↗
NASA NTRS1995

The TMI regenerable solid oxide fuel cell

Energy storage and production in space requires rugged, reliable hardware which minimizes weight, volume, and maintenance while maximizing power output and usable energy storage. These systems generally consist of photovoltaic solar arrays which operate during sunlight cycles to provide system power and regenerate fuel (hydrogen) via water electrolysis; during dark cycles, hydrogen is converted by the fuel cell into system. The currently preferred configuration uses two separate systems (fuel cell and electrolyzer) in conjunction with photovoltaic cells. Fuel cell/electrolyzer system simplicity, reliability, and power-to-weight and power-to-volume ratios could be greatly improved if both power production (fuel cell) and power storage (electrolysis) functions can be integrated into a single unit. The Technology Management, Inc. (TMI), solid oxide fuel cell-based system offers the opportunity to both integrate fuel cell and electrolyzer functions into one unit and potentially simplify system requirements. Based an the TMI solid oxide fuel cell (SOPC) technology, the TMI integrated fuel cell/electrolyzer utilizes innovative gas storage and operational concepts and operates like a rechargeable 'hydrogen-oxygen battery'. Preliminary research has been completed on improved H2/H2O electrode (SOFC anode/electrolyzer cathode) materials for solid oxide, regenerative fuel cells. Improved H2/H2O electrode materials showed improved cell performance in both fuel cell and electrolysis modes in reversible cell tests. ln reversible fuel cell/electrolyzer mode, regenerative fuel cell efficiencies (ratio of power out (fuel cell mode) to power in (electrolyzer model)) improved from 50 percent (using conventional electrode materials) to over 80 percent. The new materials will allow the TMI SOFC system to operate as both the electrolyzer and fuel cell in a single unit. Preliminary system designs have also been developed which indicate the technical feasibility of using the TMI SOFC technology for space applications with high energy storage efficiencies and high specific energy. Development of small space systems would also have potential dual-use, terrestrial applications.

Cable, Thomas L.↗
DOE OSTI2016

PGSFR Core Design and Performance Characteristics

Korea Atomic Energy Research Institute (KAERI) has developed a Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) as a representative of Gen-IV concepts. The final PGSFR core will be composed of transuranic (TRU) fuels to transmute LWR spent fuels and self-recycled its fuels, in combination with a pyro-processing facility. However, the initial POSFR core consists of low enriched uranium (LEU) fuels in order to: 1) provide an irradiation test for TRU fuels and structural materials, 2) keep pace with the development of pyro-processing technology, and 3) obtain operational experience on the reactor systems and components. During the first phase of PGSFR development plan (2012-2015), the power level was decided to be 150 MWe from many trade-off studies using different power level. The PGSFR uranium core performance characteristics at the initial, transition and equilibrium states have been carried out. In this paper, the PGSFR core design and related performance characteristics are introduced. The PGSFR uranium core design was performed at the initial/transition/equilibrium core based on the candidate fuel management scheme. It was confirmed that this design can satisfy surely all design limit and requirement. The chapter 4 of PGSFR Preliminary Safety Information Document (PSID) published in December 2015 contains the detailed information and characteristics of PGSFR core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Development and Test Results of Thermocouples Used in the TRISO-Fuel Irradiation Experiment AGR-5/6/7

High-temperature gas reactor irradiation experiments create unique challenges for thermocouple-based temperature measurements. High-temperature industrial thermocouples suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. For lower-temperature applications, Type K and Type N thermocouples are affected by neutron irradiation only to a limited extent. But until recently, the use of these nickel-based thermocouples was limited when the temperature exceeded 1050°C due to drift related to phenomena other than nuclear irradiation. Certain portions of the AGR-5/6/7 experiment experienced temperatures higher than any of the previous AGR tests, up to 1500°C. Recognizing the limitations of existing thermometry to measure such high temperatures, the sponsor of the AGR-5/6/7 test supported a development and testing program for thermocouples capable of low-drift operation at temperatures above 1100°C. This program included additional development of high-temperature irradiation-resistant thermocouples based on molybdenum/niobium thermoelements, which have been studied at Idaho National Laboratory since circa 2004. A step change in accuracy and long-term stability of this thermocouple type was achieved as part of the AGR-5/6/7 thermometry development program. Additionally, long-term testing (7000+ h) at 1250°C of Type N thermocouples utilizing a customized sheath developed at the University of Cambridge has been completed with excellent low-drift results. The results of this testing as well as testing of the improved high-temperature irradiation-resistant design are reported herein. Both the improved high-temperature irradiation-resistant and the Cambridge Type N thermocouple types were incorporated into the AGR-5/6/7 test, which began irradiation in February 2018 and was completed in July 2020. Finally, a summary of the performance of the thermocouples incorporated into the AGR-5/6/7 test is included herein.

In-pile instrumentation↗
DOE OSTI2026

EDS Analysis of FCCI in AFC-FAST Fuel Pins

The Advanced Fuel Campaign’s Fission Accelerated Steady-state Testing (FAST) program uses metallic fuel pins with small diameters to reach a desired burnup more quickly. This enables accelerated testing of advanced fuel designs and decreases the time between idea conception and commercial usage. Post-irradiation examination is critical in this process, especially with respect to the fuel-cladding chemical interactions (FCCI). In this work, energy dispersive X-ray spectroscopy (EDS) is used to track how elements from the fuel and fission products have diffused through the cladding. Elemental redistribution along the fuel-cladding interface is mapped and FCCI region thicknesses are measured. The correlations between geometry, temperature, burnup, and FCCI thickness are presented.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Experimental test results of a generalized parameter fuel control

Considerable interest has been generated recently in low cost jet propulsion systems. One of the more complicated components of jet engines is the fuel control. Results of an effort to develop a simpler hydromechanical fuel control are presented. This prototype fuel control was installed on a J85-GE-13 jet engine. Results show that the fuel control provided satisfactory engine performance at sea level static conditions over its normal nonafterburning operating range, including startup. Results of both bench and engine tests are presented; the difficulties encountered are described.

Batterton, P. G.↗