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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 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Evaporation and propagation of liquid drop streams at vacuum pressures: Experiments and modeling

Evaporation of streams of liquid droplets in environments at vacuum pressures below the vapor pressure has not been widely studied. Here, experiments and simulations are reported that quantify the change in droplet diameter when a steady stream of ≈100 μm diameter drops is injected into a chamber initially evacuated to <10 –8 bar. In experiments, droplets fall through the center of a 0.8 m long liquid nitrogen cooled shroud, simulating infinity radiation and vapor mass flux boundary conditions. Furthermore, experimentally measured changes in drop diameters vary from ≈0 to 6 μm when the initial vapor pressure is increased from 10 –6 to 10 –3 bar by heating the liquid. Measured diameter changes are predicted by a model based on the Hertz-Knudsen equation. One uncertainty in the calculation is the “sticking coefficient” β. Assuming a constant β for all conditions studied here, predicted diameter changes best match measurements with β≈0.3. This value falls within the range of β reported in the literature for organic liquids. Finally, at the higher vapor pressure conditions considered here, the drop stream disperses transverse to the main flow direction. This spread is attributed to forces imparted by an absolute pressure gradient produced by the evaporating stream.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2024

Design and Cooling Performance of Additively Manufactured Ceramic Turbine Vanes

Ceramic materials are of significant interest in aviation and power generation gas turbine engines due to their low density and ability to withstand high temperatures. Increased cycle thermal efficiency and higher specific power output is possible by incorporating ceramic components that enable high turbine inlet temperatures and lower required cooling airflow levels. However, ceramics can be difficult and costly to form into the complex shapes used in gas turbine components, often requiring specialized multi-step processes. Furthermore, ceramic components in the hottest areas of a gas turbine, such as vanes or blade shroud seals, will still likely require cooling which is challenging to implement in conventional ceramic manufacturing approaches. Therefore, this study presents a multidisciplinary approach that investigates the design, fabrication, and overall cooling effectiveness evaluation of additively manufactured (AM), polymer derived ceramic (PDC) turbine vanes. A thermo-mechanically optimized vane design was generated, ceramic additive manufacturing of the complex cooling configuration was developed, and quantification of the increase in overall cooling effectiveness was performed in a 1X scale, high-speed facility using infrared thermography. This study produced a PDC AM process, capable of printing complex internal cooling schemes in 1X scale turbine vanes. Furthermore, it was found that the optimized vane more than doubled the overall cooling effectiveness observed in the baseline design, which reasonably agreed with thermomechanical optimization model predictions. Additionally, the optimized ceramic vane outperformed an identical metal vane, in terms of area averaged cooling effectiveness, suggesting that the ceramic vane could operate at reduced coolant flowrates to achieve comparable levels of cooling performance.

Reynolds number↗
DOE OSTI2023

Design and Cooling Performance of Additively Manufactured Ceramic Turbine Vanes

Ceramic materials are of significant interest in aviation and power generation gas turbine engines due to their low density and ability to withstand high temperatures. Increased cycle thermal efficiency and higher specific power output is possible by incorporating ceramic components that enable high turbine inlet temperatures and lower required cooling airflow levels. However, ceramics can be difficult and costly to form into the complex shapes used in gas turbine components, often requiring specialized multi-step processes. Furthermore, ceramic components in the hottest areas of a gas turbine, such as vanes or blade shroud seals, will still likely require cooling which is challenging to implement in conventional ceramic manufacturing approaches. Therefore, this study presents a multidisciplinary approach that investigates the design, fabrication and overall cooling effectiveness evaluation of additively manufactured (AM), polymer derived ceramic (PDC) turbine vanes. A thermo-mechanically optimized vane design was generated, ceramic additive manufacturing of the complex cooling configuration was developed, and quantification of the increase in overall cooling effectiveness was performed in a 1X scale, high-speed facility using infrared thermography. This study produced a PDC AM process, capable of printing complex internal cooling schemes in 1X scale turbine vanes. It was found that the optimized vane more than doubled the overall cooling effectiveness observed in the baseline design, which reasonably agreed with thermomechanical optimization model predictions. Additionally, the optimized ceramic vane outperformed an identical metal vane, in terms of area averaged cooling effectiveness, suggesting that the ceramic vane could operate at reduced coolant flow rates to achieve comparable levels of cooling performance.

Fox, Andrew↗
DOE OSTI2025

Experimental Analysis of Advanced Turbine for Supercritical CO2 Power Cycle

An advanced first stage high pressure turbine blade for a supercritical CO2 power cycle is tested in the Big Rig for Aerothermal Stationary Turbine Analysis (BRASTA) annular cascade at the Purdue Experimental Turbine Aerothermal Laboratory (PETAL) high pressure blow down facility alongside a baseline blade for comparison of aerodynamic performance. Both geometries are tested simultaneously using a novel off-axis rig design to allow for the blade geometries to be scaled up to achieve higher Reynolds numbers. The off-axis design necessitates the use of discrete sectors of airfoils, which are designed and additively manufactured in house using an mSLA printer. Printing the blades allows for unique routing of passage for blade surface static pressure taps that contour to the blade shape instead of requiring a straight line view from surface tap to egress, allowing all instrumentation for 15%, 50%, and 85% span to exist in the same passage. A flow conditioning gauze [1] is placed upstream of the blade passages to impart pressure, Mach, and swirl profiles to mimic the rotor relative frame inlet conditions to the blade. Downstream, a sonic valve is used to alter the backpressure to achieve different pressure ratios for different blowdown setpoints. The design of the rig, sonic wheel, and instrumentation are discussed along with the manufacturing and GD&T of the blades with respect to the build-up of this novel experimental apparatus. Initial commissioning data is obtained from inlet total pressure and temperature rakes, blade static pressure taps, and exit total pressure rakes. Oil visualization is performed using a silicone oil mixture containing titanium dioxide and pigmented for contrast of the blade suction side surface with the hub and shroud endwalls. The viscosity of the oil is tuned so that the oil does not fully thin and blow away during the time of the test, and endoscopic cameras are placed in the rig for live monitoring and recording. The preliminary oil visualization and rake profiles are presented, showcasing the operability of this additively manufactured off-axis flow path design.

Tuite, Logan [Purdue University]↗
DOE OSTI2016

SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vessels – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. To use Grizzly for probabilistic analysis, it is necessary to have a way to rapidly evaluate stress intensity factors. To accomplish this goal, a reduced order model (ROM) has been developed to efficiently represent the behavior of a detailed 3D Grizzly model used to calculate fracture parameters. This approach uses the stress intensity factor influence coefficient method that has been used with great success in FAVOR. Instead of interpolating between tabulated solutions, as FAVOR does, the ROM approach uses a response surface methodology to compute fracture solutions based on a sampled set of results used to train the ROM. The main advantages of this approach are that the process of generating the training data can be fully automated, and the procedure can be readily used to consider more general flaw configurations. This paper demonstrates the procedure used to generate a ROM to rapidly compute stress intensity factors for axis-aligned flaws. The results from this procedure are in good agreement with those produced using the traditional influence coefficient interpolation procedure, which gives confidence in this method. This paves the way for applying this procedure for more general flaw configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

TU-F-CAMPUS-I-04: Head-Only Asymmetric Gradient System Evaluation: ACR Image Quality and Acoustic Noise

Purpose: A high-performance head-only magnetic resonance imaging gradient system with an acquisition volume of 26 cm employing an asymmetric design for the transverse coils has been developed. It is able to reach a magnitude of 85 mT/m at a slew rate of 700 T/m/s, but operated at 80 mT/m and 500 T/m/s for this test. A challenge resulting from this asymmetric design is that the gradient nonlinearly exhibits both odd- and even-ordered terms, and as the full imaging field of view is often used, the nonlinearity is pronounced. The purpose of this work is to show the system can produce clinically useful images after an on-site gradient nonlinearity calibration and correction, and show that acoustic noise levels fall within non-significant risk (NSR) limits for standard clinical pulse sequences. Methods: The head-only gradient system was inserted into a standard 3T wide-bore scanner without acoustic damping. The ACR phantom was scanned in an 8-channel receive-only head coil and the standard American College of Radiology (ACR) MRI quality control (QC) test was performed. Acoustic noise levels were measured for several standard pulse sequences. Results: Images acquired with the head-only gradient system passed all ACR MR image quality tests; Both even and odd-order gradient distortion correction terms were required for the asymmetric gradients to pass. Acoustic noise measurements were within FDA NSR guidelines of 99 dBA (with assumed 20 dBA hearing protection) A-weighted and 140 dB for peak for all but one sequence. Note the gradient system was installed without any shroud or acoustic batting. We expect final system integration to greatly reduce noise experienced by the patient. Conclusion: A high-performance head-only asymmetric gradient system operating at 80 mT/m and 500 T/m/s conforms to FDA acoustic noise limits in all but one case, and passes all the ACR MR image quality control tests. This work was supported in part by the NIH grant 5R01EB010065.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2023

Fuel-air mixing in motored CFR engine at research octane number (RON) relevant condition

This paper presents a three-dimensional (3-D) computational fluid dynamics (CFD) study of a motored cooperative fuel research (CFR) engine at research octane number (RON) relevant condition. The boundary conditions for 3-D simulations were generated with a one-dimensional GT-Power model. For the first time in literature, a carburetor was added to a virtual CFR engine model with 3-D CFD. Therefore, the proposed setup can simulate the fuel and thermal stratifications inside the engine cylinder with realistic detail. The transient simulations in this work were performed within the Reynolds-averaged Navier-Stokes (RANS) framework with a Realizable k-ε turbulence model. Major conclusions from the present work are: (1) The in-cylinder flow of the CFR engine is swirl-dominated due to the existence of the intake valve shroud. (2) There is a significant amount of liquid droplets entering the cylinder during the intake stroke. The maximum instantaneous amount of liquid for 50% PRF 87 (containing 87% iso-octane and 13% n-heptane (v/v)) and 50% ethanol mixture is indicated to be around 26% of total injected fuel mass. (3) The heat of vaporization (HoV) of the fuel is responsible for creating both temperature and charge stratification inside the cylinder.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Decay Dose Shielding Analysis with Hybrid Unstructured Mesh/Constructive Solid Geometry Monte Carlo Calculation and ADVANTG Acceleration

The Oak Ridge National Laboratory (ORNL) Second Target Station (STS) neutron production facility is an accelerator driven pulsed neutron source that is currently being actively developed at ORNL. The neutrons are produced by proton-induced spallation reactions. A proton beam of 700 kW power is delivered to a spallation target in short, less than 1 µs long pulses, with 15 Hz repetition rate. The spallation target of ORNL STS is a rotating water-cooled tungsten target with tantalum cladding housed in a stainless-steel shroud. It is divided into 21 segments. These segments become highly activated due to spallation reactions or nuclei transmutation by the emitted neutrons. The radioactive nuclides continue to decay after ceasing operation. The decay dose rates generated from the target segments once they are removed from their operational location within the core vessel must be accurately quantified to determine the shielding configurations of remote handling tools and transport casks and to aid in planning maintenance events. To determine the shielding configurations needed for an activated target segment after ceasing operation, both the hybrid unstructured mesh (UM)/constructive solid geometry (CSG) approach that was previously utilized for STS analyses [1] and the ADVANTG code [2] were used. Even though the ADVANTG code does not include UM capability, the utilization of its advanced variance reduction technique was crucial to accelerate the extremely difficult final photon transport calculation in this analysis. This paper also describes the procedures taken to mitigate the convergence issues that occur when ADVANTG uses a source definition that does not match the source of the final Monte Carlo (MC) calculation. These convergence issues often occur because of inconsistencies between source and transport biasing parameters.

Ibrahim, Ahmad↗
DOE OSTI2016

Complex ancestries of lager-brewing hybrids were shaped by standing variation in the wild yeast Saccharomyces eubayanus

Lager-style beers constitute the vast majority of the beer market, and yet, the genetic origin of the yeast strains that brew them has been shrouded in mystery and controversy. Unlike alestyle beers, which are generally brewed with Saccharomyces cerevisiae , lagers are brewed at colder temperatures with allopolyploid hybrids of Saccharomyces eubayanus x S. cerevisiae . Since the discovery of S. eubayanus in 2011, additional strains have been isolated from South America, North America, Australasia, and Asia, but only interspecies hybrids have been isolated in Europe. Here, using genome sequence data, we examine the relationships of these wild S. eubayanus strains to each other and to domesticated lager strains. Our results support the existence of a relatively low-diversity (π = 0.00197) lineage of S. eubayanus whose distribution stretches across the Holarctic ecozone and includes wild isolates from Tibet, new wild isolates from North America, and the S. eubayanus parents of lager yeasts. This Holarctic lineage is closely related to a population with higher diversity (π = 0.00275) that has been found primarily in South America but includes some widely distributed isolates. A second diverse South American population (π = 0.00354) and two early-diverging Asian subspecies are more distantly related. We further show that no single wild strain from the Holarctic lineage is the sole closest relative of lager yeasts. Instead, different parts of the genome portray different phylogenetic signals and ancestry, likely due to outcrossing and incomplete lineage sorting. Indeed, standing genetic variation within this wild Holarctic lineage of S. eubayanus is responsible for genetic variation still segregating among modern lager-brewing hybrids. Furthermore, we conclude that the relationships among wild strains of S. eubayanus and their domesticated hybrids reflect complex biogeographical and genetic processes.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2012

Baseline Fracture Toughness and SCC Testing of Alloys X750 and XM-19 (Phase I)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) based at INL and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials. The intent of this Phase I research program is to determine baseline properties for the materials of interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.

Jackson, John H.↗
DOE OSTI2012

Surface Anchoring of Nematic Phase on Carbon Nanotubes: Nanostructure of Ultra-High Temperature Materials

Nuclear energy is a dependable and economical source of electricity. Because fuel supply sources are available domestically, nuclear energy can be a strong domestic industry that can reduce dependence on foreign energy sources. Commercial nuclear power plants have extensive security measures to protect the facility from intruders [1]. However, additional research efforts are needed to increase the inherent process safety of nuclear energy plants to protect the public in the event of a reactor malfunction. The next generation nuclear plant (NGNP) is envisioned to utilize a very high temperature reactor (VHTR) design with an operating temperature of 650-1000°C [2]. One of the most important safety design requirements for this reactor is that it must be inherently safe, i.e., the reactor must shut down safely in the event that the coolant flow is interrupted [2]. This next-generation Gen IV reactor must operate in an inherently safe mode where the off-normal temperatures may reach 1500°C due to coolant-flow interruption. Metallic alloys used currently in reactor internals will melt at such temperatures. Structural materials that will not melt at such ultra-high temperatures are carbon/graphtic fibers and carbon-matrix composites. Graphite does not have a measurable melting point; it is known to sublime starting about 3300°C. However, neutron radiation-damage effects on carbon fibers are poorly understood. Therefore, the goal of this project is to obtain a fundamental understanding of the role of nanotexture on the properties of resulting carbon fibers and their neutron-damage characteristics. Although polygranular graphite has been used in nuclear environment for almost fifty years, it is not suitable for structural applications because it do not possess adequate strength, stiffness, or toughness that is required of structural components such as reaction control-rods, upper plenum shroud, and lower core-support plate [2,3]. For structural purposes, composites consisting of strong carbon fibers embedded in a carbon matrix are needed. Such carbon/carbon (C/C) composites have been used in aerospace industry to produce missile nose cones, space shuttle leading edge, and aircraft brake-pads. However, radiation-tolerance of such materials is not adequately known because only limited radiation studies have been performed on C/C composites, which suggest that pitch-based carbon fibers have better dimensional stability than that of polyacrylonitrile (PAN) based fibers [4]. The thermodynamically-stable state of graphitic crystalline packing of carbon atoms derived from mesophase pitch leads to a greater stability during neutron irradiation [5]. The specific objectives of this project were: (i) to generating novel carbonaceous nanostructures, (ii) measure extent of graphitic crystallinity and the extent of anisotropy, and (iii) collaborate with the Carbon Materials group at Oak Ridge National Lab to have neutron irradiation studies and post-irradiation examinations conducted on the carbon fibers produced in this research project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials of interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Summary of OEM Idling Recommendations from Vehicle Owner's Manuals

The project upon which this report is based was conceived in 2012 during discussions between the East Tennessee Clean Fuels Coalition (ETCleanFuels) and Oak Ridge National Laboratory (ORNL) who both noted that a detailed summary of idling recommendations for a wide variety of engines and vehicles were not available in the literature. The two organizations agreed that ETCleanFuels would develop a first-of-its-kind collection of idling recommendations from the owner’s manuals of modern production vehicles. Vehicle engine idling, a subject that has long been debated, is largely shrouded in misinformation. The justifications for idling seem to be many: driver comfort, waiting in lines, and talking on cell phones to name a few. Assuredly, a great number of people idle because of the myths and misinformation surrounding this issue. This report addresses these myths by turning to statements taken directly from the automobile and engine manufacturers themselves.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2017

Trak Investigation of Focusing Electrode Geometries for the DARHT Axis-I Diode

An investigation was carried out on the effects of different cathode shroud geometries of the DARHT Axis-1 diode using the Trak ray tracing software. Pierce angles of 20, 30, 45, 60, and 67.5 degrees were investigated. For each geometry the current density with respect to radial position will be presented as it evolves in the longitudinal direction. In addition the emittances for each geometry are compared and this information is used to determine the optimal geometry from the selected angles. These results are compared to the baseline geometry currently employed at DARHT of a simple 2.5mm recessed velvet cathode. Of the selected angles it was found that 45 degrees produced the lowest normalized emittance value, whereas 60 degrees produced the most uniform current density profile at 1cm away from the emission surface. For the purpose of this investigation the effects of the bucking coil and solenoid around the hollow anode of the DARHT Axis I injector are neglected.

47 OTHER INSTRUMENTATION↗
DOE OSTI2015

Correcting Cross-polarization Monopulse Response of Reflector Antennas

The monopulse response of radar systems utilizing a short-focal-length offset-fed parabolic reflector can be compromised by depolarization of the signal by the target and by multipath scattering from nearby objects. The polarimetric behavior of this type of antenna is examined. The use of a shroud to reduce multipath interaction with nearby objects is also described. The mechanism through which man-made targets can introduce cross-polarization components into the scattered field is explained. Two kinds of polarization filters, suitable for linear polarization, are described for mitigating the effects of depolarization due to cross-polarization scattering. The benefit of the application of a polarization filter is demonstrated by modeling a monopulse radar system viewing a dihedral corner reflector. The model demonstrates dramatic performance improvement when the filter is used, showing that usable performance can be achieved even when the target depolarization is so severe that the cross-polarized signal is more than an order of magnitude stronger than the desired co-polarized signal. Relevant and useful reference material is also included in the form of appendices describing the relationship between different polarization representations and demonstrating the conditions under which Maxwell's equations can be considered to be scale-invariant.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report

The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach to create a temperature profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. This requires, necessarily, that the power generation profile inside of test fuel rodlets also be nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by both supporting analyses and by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The analyses and comparisons presented in this report show that ATR irradiations performed using cadmium shrouding are sufficiently prototypic that they can be used with confidence in the development and testing of fast reactor fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Simulation of QUENCH-15 and Preliminary Pre-Test Predictions for QUENCH-19

Iron Chromium Aluminum alloys (FeCrAl) have been developed at Oak Ridge National Laboratory as an advanced fuel cladding for light water reactors (LWR). In contrast to zirconium-based alloys employed today, FeCrAl has slower oxidation kinetics and higher strength. A test (QUENCH-19) at Karlsruhe Institute of Technology’s QUENCH facility is planned to test and demonstrate the performance of FeCrAl cladding under postulated accident conditions. This test will replicate the previous QUENCH-15 test but will use FeCrAl instead of zirconium-based materials. During these tests, a small bundle of fuel rod simulators is electrically heated in a steam and argon environment. The temperature history, oxidation thickness, and in-situ hydrogen generation are the primary measurements of interest.The CORA/BWR code was previously developed and utilized in the 1980’s and 1990’s to simulate BWR bundle degradation during the CORA test series. A modified version of the CORA/BWR code, named CORA/BWR-QUENCH, has been developed to model the QUENCH test facility. The CORA/BWR-QUENCH code was used to model the QUENCH-15 test which used zirconium-based materials. The code predictions and test data were found to be in good agreement. The code was then used to model the planned QUENCH-19 test utilizing FeCrAl cladding, shroud, and spacer grids. The simulation predicts the peak cladding temperature to be 1620 K (~1347°C) for QUENCH-19 which is significantly lower than that during QUENCH-15 with the zirconium-based alloy. The predicted hydrogen generation is also significantly lower for the FeCrAl clad QUENCH-19 as compared to QUENCH-15: ~1.7 g versus ~48.8 g. These results align with expectations given the much slower oxidation kinetics of FeCrAl alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗