Studies of behavior of the fuel compound based on the U-Zr micro-heterogeneous quasialloy during cyclic thermal tests

NASA Astrophysics Data System (ADS)

Zaytsev, D. A.; Repnikov, V. M.; Soldatkin, D. M.; Solntsev, V. A.

2017-11-01

This paper provides the description of temperature cycle testing of U-Zr heterogeneous fuel composition. The composition is essentially a niobium-doped zirconium matrix with metallic uranium filaments evenly distributed over the cross section. The test samples 150 mm long had been fabricated using a fiber-filament technology. The samples were essentially two-bladed spiral mandrel fuel elements parts. In the course of experiments the following temperatures were applied: 350, 675, 780 and 1140 °C with total exposure periods equal to 200, 30, 30 and 6 hours respectively. The fuel element samples underwent post-exposure material science examination including: geometry measurements, metallographic analysis, X-ray phase analysis and electron-microscopic analysis as well as micro-hardness measurement. It has been found that no significant thermal swelling of the samples occurs throughout the whole temperature range from 350 °C up to 1140 °C. The paper presents the structural changes and redistribution of the fuel component over the fuel element cross section with rising temperature.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

2013-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

Investigations of a Combustor Using a 9-Point Swirl-Venturi Fuel Injector: Recent Experimental Results

NASA Technical Reports Server (NTRS)

Hicks, Yolanda R.; Heath, Christopher M.; Anderson, Robert C.; Tacina, Kathleen M.

2012-01-01

This paper explores recent results obtained during testing in an optically-accessible, JP8-fueled, flame tube combustor using baseline Lean Direct Injection (LDI) research hardware. The baseline LDI geometry has nine fuel/air mixers arranged in a 3 x 3 array. Results from this nine-element array include images of fuel and OH speciation via Planar Laser-Induced Fluorescence (PLIF), which describe fuel spray pattern and reaction zones. Preliminary combustion temperatures derived from Stokes/Anti-Stokes Spontaneous Raman Spectroscopy are also presented. Other results using chemiluminescence from major combustion radicals such as CH* and C2* serve to identify the primary reaction zone, while OH PLIF shows the extent of reaction further downstream. Air and fuel velocities and fuel drop size results are also reported.

Axisymmetric single shear element combustion instability experiment

NASA Technical Reports Server (NTRS)

Breisacher, Kevin J.

1993-01-01

The combustion stability characteristics of a combustor consisting of a single shear element and a cylindrical chamber utilizing LOX and gaseous hydrogen as propellants are presented. The combustor geometry and the resulting longitudinal mode instability are axisymmetric. Hydrogen injection temperature and pyrotechnic pulsing were used to determine stability boundaries. Mixture ratio, fuel annulus gap, and LOX post configuration were varied. Performance and stability data were obtained for chamber pressures of 300 and 1000 psia.

Axisymmetric single shear element combustion instability experiment

NASA Technical Reports Server (NTRS)

Breisacher, Kevin J.

1993-01-01

The combustion stability characteristics of a combustor consisting of a single shear element and a cylindrical chamber utilizing LOX and gaseous hydrogen as propellants are presented. The combustor geometry and the resulting longitudinal mode instability are axisymmetric. Hydrogen injection temperature and pyrotechnic pulsing were used to determine stability boundaries. Mixture ratio, fuel annulus gap, and LOX post configuration were varied. Performance and stability data are presented for chamber pressures of 300 and 1000 psia.

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonzalez-Pintor, S.; Verdu, G.; Ginestar, D.

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmarkmore » and with the results provided by PARCS code. (authors)« less

Numerical Optimisation in Non Reacting Conditions of the Injector Geometry for a Continuous Detonation Wave Rocket Engine

NASA Astrophysics Data System (ADS)

Gaillard, T.; Davidenko, D.; Dupoirieux, F.

2015-06-01

The paper presents the methodology and the results of a numerical study, which is aimed at the investigation and optimisation of different means of fuel and oxidizer injection adapted to rocket engines operating in the rotating detonation mode. As the simulations are achieved at the local scale of a single injection element, only one periodic pattern of the whole geometry can be calculated so that the travelling detonation waves and the associated chemical reactions can not be taken into account. Here, separate injection of fuel and oxidizer is considered because premixed injection is handicapped by the risk of upstream propagation of the detonation wave. Different associations of geometrical periodicity and symmetry are investigated for the injection elements distributed over the injector head. To analyse the injection and mixing processes, a nonreacting 3D flow is simulated using the LES approach. Performance of the studied configurations is analysed using the results on instantaneous and mean flowfields as well as by comparing the mixing efficiency and the total pressure recovery evaluated for different configurations.

RELAP5 Application to Accident Analysis of the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less

Local Heat Flux Measurements with Single Element Coaxial Injectors

NASA Technical Reports Server (NTRS)

Jones, Gregg; Protz, Christopher; Bullard, Brad; Hulka, James

2006-01-01

To support the mission for the NASA Vision for Space Exploration, the NASA Marshall Space Flight Center conducted a program in 2005 to improve the capability to predict local thermal compatibility and heat transfer in liquid propellant rocket engine combustion devices. The ultimate objective was to predict and hence reduce the local peak heat flux due to injector design, resulting in a significant improvement in overall engine reliability and durability. Such analyses are applicable to combustion devices in booster, upper stage, and in-space engines, as well as for small thrusters with few elements in the injector. In this program, single element and three-element injectors were hot-fire tested with liquid oxygen and ambient temperature gaseous hydrogen propellants at The Pennsylvania State University Cryogenic Combustor Laboratory from May to August 2005. Local heat fluxes were measured in a 1-inch internal diameter heat sink combustion chamber using Medtherm coaxial thermocouples and Gardon heat flux gauges. Injectors were tested with shear coaxial and swirl coaxial elements, including recessed, flush and scarfed oxidizer post configurations, and concentric and non-concentric fuel annuli. This paper includes general descriptions of the experimental hardware, instrumentation, and results of the hot-fire testing for three of the single element injectors - recessed-post shear coaxial with concentric fuel, flush-post swirl coaxial with concentric fuel, and scarfed-post swirl coaxial with concentric fuel. Detailed geometry and test results will be published elsewhere to provide well-defined data sets for injector development and model validatation.

Study of the Optimum Zone of the Independent Variables of an ORGEL Reactor Connected to a 250-MWeb Power Plant. Self Supporting Fuel Elements Made of UC, with Sap Cladding with Four Fuel Rods and Individual Pressure Tubes; STUDIE DER OPTIMALEN ZONE DER UNABHANGIGEN PARAMETER EINES ORGEL- REAKTORS IN EINEM 250-MWe-KRAFTWERK. SELBST-TRAGENDES BRENNELEMENT AUS UC, SAP-UMHUL-LUNG MIT 4 BRENNSTOFFSTABEN UND INDIVIDUELLEN DRUCKROHREN

DOE Office of Scientific and Technical Information (OSTI.GOV)

LaFontaine, F.; Tauch, P.

The optimum range of the independent variables of and ORGEL reactor connected to a 250-Mw power plant (4 fuel rods of UC with individual pressure tubes), as well as the geometry of the reactor core and the operation of the plant, is described. (auth)

Study for the optimization of a transport aircraft wing for maximum fuel efficiency. Volume 1: Methodology, criteria, aeroelastic model definition and results

NASA Technical Reports Server (NTRS)

Radovcich, N. A.; Dreim, D.; Okeefe, D. A.; Linner, L.; Pathak, S. K.; Reaser, J. S.; Richardson, D.; Sweers, J.; Conner, F.

1985-01-01

Work performed in the design of a transport aircraft wing for maximum fuel efficiency is documented with emphasis on design criteria, design methodology, and three design configurations. The design database includes complete finite element model description, sizing data, geometry data, loads data, and inertial data. A design process which satisfies the economics and practical aspects of a real design is illustrated. The cooperative study relationship between the contractor and NASA during the course of the contract is also discussed.

Study of Compton suppression for use in spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Bender, Sarah

The focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any additional peaks, due to its relatively low resolution. Samples of Approved Test Material (ATM) commercial fuel elements were obtained from Pacific Northwest National Laboratory. The samples had been processed using the beginning stages of the PUREX method and represented the unseparated feed solution from a reprocessing facility. Compton suppressed measurements of the ATM fuel samples were recorded inside the guard detector annulus, to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Photopeak losses were observed in the measurements of the dissolved ATM fuel samples because the spectra was recorded from the source in very close proximity to the detector and surrounded by the guard annulus, so the detection probability is very high. Though this configuration is optimal for a Compton suppression system for the measurement of low count rate samples, measurement of high count rate samples in the enclosed arrangement leads to sum peaks in both the suppressed and unsuppressed spectra and losses to photopeak counts in the suppressed spectra. No additional photopeaks were detected using Compton suppression with this geometry. A detector model was constructed that can accurately simulate a Compton suppressed spectral measurement of radiation from spent nuclear fuel using HPGe or LaBr3 detectors. This is the first detector model capable of such an accomplishment. The model uses the Geant4 toolkit coupled with the RadSrc application and it accepts spent fuel composition data in list form. The model has been validated using dissolved ATM fuel samples in the standard, enclosed geometry of the PSU HPGe-CSS. The model showed generally good agreement with both the unsuppressed and suppressed measured fuel sample spectra, however the simulation is more appropriate for the generation of gamma-ray spectra in the beam source configuration. Photopeak losses due to cascade decay emissions in the Compton suppressed spectra were not appropriately managed by the simulation. Compton suppression would be a beneficial addition to NDA process monitoring systems if oriented such that the gamma-ray photons are collimated to impinge the primary detector face as a beam. The analysis has shown that peak losses through accidental coincidences are minimal and the reduction in the Compton continuum allows additional peaks to be resolved. (Abstract shortened by UMI.).

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

NASA Astrophysics Data System (ADS)

Schneider, E. A.; Deinert, M. R.; Cady, K. B.

2006-10-01

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

Multidimensional Multiphysics Simulation of TRISO Particle Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. D. Hales; R. L. Williamson; S. R. Novascone

2013-11-01

Multidimensional multiphysics analysis of TRISO-coated particle fuel using the BISON finite-element based nuclear fuels code is described. The governing equations and material models applicable to particle fuel and implemented in BISON are outlined. Code verification based on a recent IAEA benchmarking exercise is described, and excellant comparisons are reported. Multiple TRISO-coated particles of increasing geometric complexity are considered. It is shown that the code's ability to perform large-scale parallel computations permits application to complex 3D phenomena while very efficient solutions for either 1D spherically symmetric or 2D axisymmetric geometries are straightforward. Additionally, the flexibility to easily include new physical andmore » material models and uncomplicated ability to couple to lower length scale simulations makes BISON a powerful tool for simulation of coated-particle fuel. Future code development activities and potential applications are identified.« less

Grid-to-rod flow-induced impact study for PWR fuel in reactor

DOE PAGES

Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

2016-06-10

The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

Investigation of a Tricarbide Grooved Ring Fuel Element for a Nuclear Thermal Rocket

NASA Technical Reports Server (NTRS)

Taylor, Brian D.; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2017-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the temperature limitations of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment.

Analytical and experimental investigation of rubbing interaction in labyrinth seals for a liquid hydrogen fuel pump. [space shuttle main engine

NASA Technical Reports Server (NTRS)

Dolan, F. X.; Kennedy, F. E.; Schulson, E. M.

1984-01-01

Cracking of the titanium knife edges on the labyrinth seals of the liquid hydrogen fuel pump in the Space Shuttle main engine is considered. Finite element analysis of the thermal response of the knife edge in sliding contact with the wear ring surface shows that interfacial temperatures can be quite high and they are significantly influenced by the thermal conductivity of the surfaces in rubbing contact. Thermal shock experiments on a test specimen similar to the knife edge geometry demonstrate that cracking of the titanium alloy is possible in a situation involving repeated thermal cycles over a wide temperature range, as might be realized during a rub in the liquid hydrogen fuel pump. High-speed rub interaction tests were conducted using a representative knife edge and seal geometry over a broad range of interaction rates and alternate materials were experimentally evaluated. Plasma-sprayed aluminum-graphite was found to be significantly better than presently used aluminum alloy seals from the standpoint of rub performance. Ion nitriding the titanium alloy knife-edges also improved rub performance compared to the untreated baseline.

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hales, J. D.; Williamson, R. L.; Novascone, S. R.

2016-09-01

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact,more » and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.« less

Crash Simulation of a Vertical Drop Test of a Commuter-Class Aircraft

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Fasanella, Edwin L.

2004-01-01

A finite element model of an ATR42-300 commuter-class aircraft was developed and a crash simulation was executed. Analytical predictions were correlated with data obtained from a 30-ft/s (9.14-m/s) vertical drop test of the aircraft. The purpose of the test was to evaluate the structural response of the aircraft when subjected to a severe, but survivable, impact. The aircraft was configured with seats, dummies, luggage, and other ballast. The wings were filled with 8,700 lb. (3,946 kg) of water to represent the fuel. The finite element model, which consisted of 57,643 nodes and 62,979 elements, was developed from direct measurements of the airframe geometry. The seats, dummies, luggage, fuel, and other ballast were represented using concentrated masses. The model was executed in LS-DYNA, a commercial code for performing explicit transient dynamic simulations. Predictions of structural deformation and selected time-history responses were generated. The simulation was successfully validated through extensive test-analysis correlation.

Test-Analysis Correlation of a Crash Simulation of a Vertical Drop Test of a Commuter-Category Aircraft

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Fasanella, Edwin L.

2004-01-01

A finite element model of an ATR42-300 commuter-class aircraft was developed and a crash simulation was executed. Analytical predictions were correlated with data obtained from a 30-feet per second (9.14-meters per second) vertical drop test of the aircraft. The purpose of the test was to evaluate the structural response of the aircraft when subjected to a severe, but survivable, impact. The aircraft was configured with seats, dummies, luggage, and other ballast. The wings were filled with 8,700 lb. (3,946 kilograms) of water to represent the fuel. The finite element model, which consisted of 57,643 nodes and 62,979 elements, was developed from direct measurements of the airframe geometry. The seats, dummies, luggage, simulated engines and fuel, and other ballast were represented using concentrated masses. The model was executed in LS-DYNA, a commercial finite element code for performing explicit transient dynamic simulations. Analytical predictions of structural deformation and selected time-history responses were correlated with experimental data from the drop test to validate the simulation.

Performance characterization of complex fuel port geometries for hybrid rocket fuel grains

NASA Astrophysics Data System (ADS)

Bath, Andrew

This research investigated the 3D printing and burning of fuel grains with complex geometry and the development of software capable of modeling and predicting the regression of a cross-section of these complex fuel grains. The software developed did predict the geometry to a fair degree of accuracy, especially when enhanced corner rounding was turned on. The model does have some drawbacks, notably being relatively slow, and does not perfectly predict the regression. If corner rounding is turned off, however, the model does become much faster; although less accurate, this method does still predict a relatively accurate resulting burn geometry, and is fast enough to be used for performance-tuning or genetic algorithms. In addition to the modeling method, preliminary investigations into the burning behavior of fuel grains with a helical flow path were performed. The helix fuel grains have a regression rate of nearly 3 times that of any other fuel grain geometry, primarily due to the enhancement of the friction coefficient between the flow and flow path.

Evaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste - 13187

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fast, Ivan; Aksyutina, Yuliya; Tietze-Jaensch, Holger

2013-07-01

Burn up calculations facilitate a determination of the composition and nuclear inventory of spent nuclear fuel, if operational history is known. In case this information is not available, the total nuclear inventory can be determined by means of destructive or, even on industrial scale, nondestructive measurement methods. For non-destructive measurements however only a few easy-to-measure, so-called key nuclides, are determined due to their characteristic gamma lines or neutron emission. From these measured activities the fuel burn up and cooling time are derived to facilitate the numerical inventory determination of spent fuel elements. Most regulatory bodies require an independent assessment ofmore » nuclear waste properties and their documentation. Prominent part of this assessment is a consistency check of inventory declaration. The waste packages often contain wastes from different types of spent fuels of different history and information about the secondary reactor parameters may not be available. In this case the so-called characteristic fuel burn up and cooling time are determined. These values are obtained from a correlations involving key-nuclides with a certain bandwidth, thus with upper and lower limits. The bandwidth is strongly dependent on secondary reactor parameter such as initial enrichment, temperature and density of the fuel and moderator, hence the reactor type, fuel element geometry and plant operation history. The purpose of our investigation is to look into the scaling and correlation limitations, to define and verify the range of validity and to scrutinize the dependencies and propagation of uncertainties that affect the waste inventory declarations and their independent verification. This is accomplished by numerical assessment and simulation of waste production using well accepted codes SCALE 6.0 and 6.1 to simulate the cooling time and burn up of a spent fuel element. The simulations are benchmarked against spent fuel from the real reactor Obrigheim in Germany for which sufficiently precise experimental reference data are available. (authors)« less

Experimental Combustion Dynamics Behavior of a Multi-Element Lean Direct Injection (LDI) Gas Turbine Combustor

NASA Technical Reports Server (NTRS)

Acosta, Waldo A.; Chang, Clarence T.

2016-01-01

An experimental investigation of the combustion dynamic characteristics of a research multi-element lean direct injection (LDI) combustor under simulated gas turbine conditions was conducted. The objective was to gain a better understanding of the physical phenomena inside a pressurized flametube combustion chamber under acoustically isolated conditions. A nine-point swirl venturi lean direct injection (SV-LDI) geometry was evaluated at inlet pressures up to 2,413 kPa and non-vitiated air temperatures up to 867 K. The equivalence ratio was varied to obtain adiabatic flame temperatures between 1388 K and 1905 K. Dynamic pressure measurements were taken upstream of the SV-LDI, in the combustion zone and downstream of the exit nozzle. The measurements showed that combustion dynamics were fairly small when the fuel was distributed uniformly and mostly due to fluid dynamics effects. Dynamic pressure fluctuations larger than 40 kPa at low frequencies were measured at 653 K inlet temperature and 1117 kPa inlet pressure when fuel was shifted and the pilot fuel injector equivalence ratio was increased to 0.72.

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Novascone, Stephen Rhead; Gardner, Russell James

2016-06-01

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation andmore » power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Design and Fabrication of Oxygen/RP-2 Multi-Element Oxidizer-Rich Staged Combustion Thrust Chamber Injectors

NASA Technical Reports Server (NTRS)

Garcia, C. P.; Medina, C. R.; Protz, C. S.; Kenny, R. J.; Kelly, G. W.; Casiano, M. J.; Hulka, J. R.; Richardson, B. R.

2016-01-01

As part of the Combustion Stability Tool Development project funded by the Air Force Space and Missile Systems Center, the NASA Marshall Space Flight Center was contracted to assemble and hot-fire test a multi-element integrated test article demonstrating combustion characteristics of an oxygen/hydrocarbon propellant oxidizer-rich staged-combustion engine thrust chamber. Such a test article simulates flow through the main injectors of oxygen/kerosene oxidizer-rich staged combustion engines such as the Russian RD-180 or NK-33 engines, or future U.S.-built engine systems such as the Aerojet-Rocketdyne AR-1 engine or the Hydrocarbon Boost program demonstration engine. On the current project, several configurations of new main injectors were considered for the thrust chamber assembly of the integrated test article. All the injector elements were of the gas-centered swirl coaxial type, similar to those used on the Russian oxidizer-rich staged-combustion rocket engines. In such elements, oxidizer-rich combustion products from the preburner/turbine exhaust flow through a straight tube, and fuel exiting from the combustion chamber and nozzle regenerative cooling circuits is injected near the exit of the oxidizer tube through tangentially oriented orifices that impart a swirl motion such that the fuel flows along the wall of the oxidizer tube in a thin film. In some elements there is an orifice at the inlet to the oxidizer tube, and in some elements there is a sleeve or "shield" inside the oxidizer tube where the fuel enters. In the current project, several variations of element geometries were created, including element size (i.e., number of elements or pattern density), the distance from the exit of the sleeve to the injector face, the width of the gap between the oxidizer tube inner wall and the outer wall of the sleeve, and excluding the sleeve entirely. This paper discusses the design rationale for each of these element variations, including hydraulic, structural, thermal, combustion performance, and combustion stability considerations. This paper also discusses the fabrication and assembly of the injector components, including the injector body/interpropellant plate, the additive manufactured GRCop-84 faceplate, and the pieces that make up the injector elements including the oxidizer tube, an inlet to the oxidizer tube, and a facenut that includes the fuel tangential inlets and forms the initial recessed volume where oxidizer and fuel first interact. Hot-fire test results of these main injector designs in an integrated test article that includes an oxidizer-rich preburner are described in companion papers at this JANNAF meeting.

Transient heat and mass transfer analysis in a porous ceria structure of a novel solar redox reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandran, RB; Bader, R; Lipinski, W

2015-06-01

Thermal transport processes are numerically analyzed for a porous ceria structure undergoing reduction in a novel redox reactor for solar thermochemical fuel production. The cylindrical reactor cavity is formed by an array of annular reactive elements comprising the porous ceria monolith integrated with gas inlet and outlet channels. Two configurations are considered, with the reactor cavity consisting of 10 and 20 reactive elements, respectively. Temperature dependent boundary heat fluxes are obtained on the irradiated cavity wall by solving for the surface radiative exchange using the net radiation method coupled to the heat and mass transfer model of the reactive element.more » Predicted oxygen production rates are in the range 40-60 mu mol s(-1) for the geometries considered. After an initial rise, the average temperature of the reactive element levels off at 1660 and 1680 K for the two geometries, respectively. For the chosen reduction reaction rate model, oxygen release continues after the temperature has leveled off which indicates that the oxygen release reaction is limited by chemical kinetics and/or mass transfer rather than by the heating rate. For a fixed total mass of ceria, the peak oxygen release rate is doubled for the cavity with 20 reactive elements due to lower local oxygen partial pressure. (C) 2015 Elsevier Masson SAS. All rights reserved.« less

Simulation of Reacting Flow with a Discontinuous Spectral Element Method

NASA Astrophysics Data System (ADS)

Ghiasi, Zia; Mashayek, Farzad; Komperda, Jonathan

2013-11-01

While using high order methods is desirable in order to accurately capture the small scale mixing effects in reacting flows, the challenge is to develop and implement such methods for complex geometries. In this work, a high-order Discontinuous Spectral Element Method (DSEM) code, which solves for the Navier-Stokes equations, has been modified by adding the appropriate components to solve for scalar transport equations in order to simulate the chemical reaction. Dealing with discontinuous solution at element interfaces is a challenge that is met by patching the fluxes at mortars thus making them continuous on interfaces. The patching is performed using the Lax-Fredrichs numerical flux for scalars, whereas a generalized Riemann solver is used for the Navier-Stokes equations. Direct numerical simulation is conducted in a temporally developing mixing layer to validate the method for a single step reaction (F + rO --> [ 1 + r ] P). Next, the method is implemented to simulate a subsonic reacting flow in a slanted cavity combustor with gaseous fuel injectors to demonstrate the capability of the method to handle complex geometries. The results will be used for physical understanding of mixing and reaction in this type of combustors.

Numerical and experimental investigation of the effect of geometry on combustion characteristics of solid-fuel ramjet

NASA Astrophysics Data System (ADS)

Gong, Lunkun; Chen, Xiong; Musa, Omer; Yang, Haitao; Zhou, Changsheng

2017-12-01

Numerical and experimental investigation on the solid-fuel ramjet was carried out to study the effect of geometry on combustion characteristics. The two-dimensional axisymmetric program developed in the present study adopted finite rate chemistry and second-order moment turbulence-chemistry models, together with k-ω shear stress transport (SST) turbulence model. Experimental data were obtained by burning cylindrical polyethylene using a connected pipe facility. The simulation results show that a fuel-rich zone near the solid fuel surface and an air-rich zone in the core exist in the chamber, and the chemical reactions occur mainly in the interface of this two regions; The physical reasons for the effect of geometry on regression rate is the variation of turbulent viscosity due to the geometry change. Port-to-inlet diameter ratio is the main parameter influencing the turbulent viscosity, and a linear relationship between port-to-inlet diameter and regression rate were obtained. The air mass flow rate and air-fuel ratio are the main influencing factors on ramjet performances. Based on the simulation results, the correlations between geometry and air-fuel ratio were obtained, and the effect of geometry on ramjet performances was analyzed according to the correlation. Three-dimensional regression rate contour obtained experimentally indicates that the regression rate which shows axisymmetric distribution due to the symmetry structure increases sharply, followed by slow decrease in axial direction. The radiation heat transfer in recirculation zone cannot be ignored. Compared with the experimental results, the deviations of calculated average regression rate and characteristic velocity are about 5%. Concerning the effect of geometry on air-fuel ratio, the deviations between experimental and theoretical results are less than 10%.

TRISO-fuel element thermo-mechanical performance modeling for the hybrid LIFE engine with Pu fuel blanket

NASA Astrophysics Data System (ADS)

DeMange, P.; Marian, J.; Caro, M.; Caro, A.

2010-10-01

A TRISO-coated fuel thermo-mechanical performance study is performed for the fusion-fission hybrid Laser Inertial Fusion Engine (LIFE) to test the viability of TRISO particles to achieve ultra-high burn-up of Pu or transuranic spent nuclear fuel blankets. Our methodology includes full elastic anisotropy, time and temperature varying material properties, and multilayer capabilities. In order to achieve fast fluences up to 30 × 10 25 n m -2 ( E > 0.18 MeV), judicious extrapolations across several orders of magnitude of existing material databases have been carried out. The results of our study indicate that failure of the pyrolytic carbon (PyC) layers occurs within the first 2 years of operation. The particles then behave as a single-SiC-layer particle and the SiC layer maintains reasonably-low tensile stresses until the end-of-life. It is also found that the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Conversely, varying the geometry of the TRISO-coated fuel particles results in little differences in terms of fuel performance.

Evaluation of Thin Plate Hydrodynamic Stability through a Combined Numerical Modeling and Experimental Effort

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tentner, A.; Bojanowski, C.; Feldman, E.

An experimental and computational effort was undertaken in order to evaluate the capability of the fluid-structure interaction (FSI) simulation tools to describe the deflection of a Missouri University Research Reactor (MURR) fuel element plate redesigned for conversion to lowenriched uranium (LEU) fuel due to hydrodynamic forces. Experiments involving both flat plates and curved plates were conducted in a water flow test loop located at the University of Missouri (MU), at conditions and geometries that can be related to the MURR LEU fuel element. A wider channel gap on one side of the test plate, and a narrower on the othermore » represent the differences that could be encountered in a MURR element due to allowed fabrication variability. The difference in the channel gaps leads to a pressure differential across the plate, leading to plate deflection. The induced plate deflection the pressure difference induces in the plate was measured at specified locations using a laser measurement technique. High fidelity 3-D simulations of the experiments were performed at MU using the computational fluid dynamics code STAR-CCM+ coupled with the structural mechanics code ABAQUS. Independent simulations of the experiments were performed at Argonne National Laboratory (ANL) using the STAR-CCM+ code and its built-in structural mechanics solver. The simulation results obtained at MU and ANL were compared with the corresponding measured plate deflections.« less

Uranium from German Nuclear Power Projects of the 1940s— A Nuclear Forensic Investigation

PubMed Central

Mayer, Klaus; Wallenius, Maria; Lützenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; van Belle, Pieter; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L Keith; Tims, Stephen G; Fröhlich, Michaela B; Steier, Peter

2015-01-01

Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel.3b,d, 4 Through measurement of the 230Th/234U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the 87Sr/86Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of 236U and 239Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. PMID:26501922

LES of cavitating flow inside a Diesel injector including dynamic needle movement

NASA Astrophysics Data System (ADS)

Örley, F.; Hickel, S.; Schmidt, S. J.; Adams, N. A.

2015-12-01

We perform large-eddy simulations (LES) of the turbulent, cavitating flow inside a 9-hole solenoid common-rail injector including jet injection into gas during a full injection cycle. The liquid fuel, vapor, and gas phases are modelled by a homogeneous mixture approach. The cavitation model is based on a thermodynamic equilibrium assumption. The geometry of the injector is represented on a Cartesian grid by a conservative cut-element immersed boundary method. The strategy allows for the simulation of complex, moving geometries with sub-cell resolution. We evaluate the effects of needle movement on the cavitation characteristics in the needle seat and tip region during opening and closing of the injector. Moreover, we study the effect of cavitation inside the injector nozzles on primary jet break-up.

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

Uranium from German Nuclear Power Projects of the 1940s--A Nuclear Forensic Investigation.

PubMed

Mayer, Klaus; Wallenius, Maria; Lützenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; van Belle, Pieter; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L Keith; Tims, Stephen G; Fröhlich, Michaela B; Steier, Peter

2015-11-02

Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel. Through measurement of the (230)Th/(234)U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the (87)Sr/(86)Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of (236)U and (239)Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been necessary to perform a careful design study of the probe geometry. For this, finite element analysis (FEA) has been performed in combination with practical validation tests on representative fuel dummies with machined flaws to find the probe geometry that best detects a hidden flaw. Tests performed thus far show that gaps down to 25 μm thickness can be detected with good repeatability and good discrimination from lift-off signals.

Medical Isotope Production Analyses In KIPT Neutron Source Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talamo, Alberto; Gohar, Yousry

Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is drivenmore » by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99Mo is the parent isotope of 99mTc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.« less

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Astrophysics Data System (ADS)

Emrich, William J.

2008-01-01

To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2008-01-01

To support the eventual development of a nuclear thermal rocket engine, a state-of-the-art experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

SMART Structures User's Guide - Version 3.0

NASA Technical Reports Server (NTRS)

Spangler, Jan L.

1996-01-01

Version 3.0 of the Solid Modeling Aerospace Research Tool (SMART Structures) is used to generate structural models for conceptual and preliminary-level aerospace designs. Features include the generation of structural elements for wings and fuselages, the integration of wing and fuselage structural assemblies, and the integration of fuselage and tail structural assemblies. The highly interactive nature of this software allows the structural engineer to move quickly from a geometry that defines a vehicle's external shape to one that has both external components and internal components which may include ribs, spars, longerons, variable depth ringframes, a floor, a keel, and fuel tanks. The geometry that is output is consistent with FEA requirements and includes integrated wing and empennage carry-through and frame attachments. This report provides a comprehensive description of SMART Structures and how to use it.

Heater Development, Fabrication, and Testing: Analysis of Fabricated Heaters

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Dickens, R. E.; Farmer, J. T.; Davis, J. D.; Adams, M. R.; Martin, J. J.; Webster, K. L.

2008-01-01

Thermal simulators (highly designed heater elements) developed at the Early Flight Fission Test Facility (EFF-TF) are used to simulate the heat from nuclear fission in a variety of reactor concepts. When inserted into the reactor geometry, the purpose of the thermal simulators is to deliver thermal power to the test article in the same fashion as if nuclear fuel were present. Considerable effort has been expended to mimic heat from fission as closely as possible. To accurately represent the fuel, the simulators should be capable of matching the overall properties of the nuclear fuel rather than simply matching the fuel temperatures. This includes matching thermal stresses in the pin, pin conductivities, total core power, and core power profile (axial and radial). This Technical Memorandum discusses the historical development of the thermal simulators used in nonnuclear testing at the EFF-TF and provides a basis for the development of the current series of thermal simulators. The status of current heater fabrication and testing is assessed, providing data and analyses for both successes and failures experienced in the heater development and testing program.

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE PAGES

Piro, M.H.A; Wassermann, F.; Grundmann, S.; ...

2017-05-23

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piro, M.H.A; Wassermann, F.; Grundmann, S.

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

DOEpatents

Weems, S.J.

1963-09-24

>A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

Fuel Regression Rate Behavior of CAMUI Hybrid Rocket

NASA Astrophysics Data System (ADS)

Kaneko, Yudai; Itoh, Mitsunori; Kakikura, Akihito; Mori, Kazuhiro; Uejima, Kenta; Nakashima, Takuji; Wakita, Masashi; Totani, Tsuyoshi; Oshima, Nobuyuki; Nagata, Harunori

A series of static firing tests was conducted to investigate the fuel regression characteristics of a Cascaded Multistage Impinging-jet (CAMUI) type hybrid rocket motor. A CAMUI type hybrid rocket uses the combination of liquid oxygen and a fuel grain made of polyethylene as a propellant. The collision distance divided by the port diameter, H/D, was varied to investigate the effect of the grain geometry on the fuel regression rate. As a result, the H/D geometry has little effect on the regression rate near the stagnation point, where the heat transfer coefficient is high. On the contrary, the fuel regression rate decreases near the circumference of the forward-end face and the backward-end face of fuel blocks. Besides the experimental approaches, a method of computational fluid dynamics clarified the heat transfer distribution on the grain surface with various H/D geometries. The calculation shows the decrease of the flow velocity due to the increase of H/D on the area where the fuel regression rate decreases with the increase of H/D. To estimate the exact fuel consumption, which is necessary to design a fuel grain, real-time measurement by an ultrasonic pulse-echo method was performed.

Monitoring arrangement for vented nuclear fuel elements

DOEpatents

Campana, Robert J.

1981-01-01

In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

Design and preliminary results of a fuel flexible industrial gas turbine combustor

NASA Technical Reports Server (NTRS)

Novick, A. S.; Troth, D. L.; Yacobucci, H. G.

1981-01-01

The design characteristics are presented of a fuel tolerant variable geometry staged air combustor using regenerative/convective cooling. The rich/quench/lean variable geometry combustor is designed to achieve low NO(x) emission from fuels containing fuel bound nitrogen. The physical size of the combustor was calculated for a can-annular combustion system with associated operating conditions for the Allison 570-K engine. Preliminary test results indicate that the concept has the potential to meet emission requirements at maximum continuous power operation. However, airflow sealing and improved fuel/air mixing are necessary to meet Department of Energy program goals.

Rolling Process Modeling Report. Finite-Element Model Validation and Parametric Study on various Rolling Process parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2015-06-15

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum alloy plate-type fuel for high-performance research reactors in the United States. This work supports the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization Reactor Conversion Program. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll-separating forces for various rolling mill geometries for PNNL, Babcock & Wilcox Co., Y-12 National Security Complex, Los Alamos National Laboratory, and Idaho National Laboratory. The model developed and presented in a previous report has been subjected to further validationmore » study using new sets of experimental data generated from a rolling mill at PNNL. Simulation results of both hot rolling and cold rolling of uranium-10% molybdenum coupons have been compared with experimental results. The model was used to predict roll-separating forces at different temperatures and reductions for five rolling mills within the National Nuclear Security Administration Fuel Fabrication Capability project. This report also presents initial results of a finite-element model microstructure-based approach to study the surface roughness at the interface between zirconium and uranium-10% molybdenum.« less

User Manual for the PROTEUS Mesh Tools

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Micheal A.; Shemon, Emily R

2016-09-19

PROTEUS is built around a finite element representation of the geometry for visualization. In addition, the PROTEUS-SN solver was built to solve the even-parity transport equation on a finite element mesh provided as input. Similarly, PROTEUS-MOC and PROTEUS-NEMO were built to apply the method of characteristics on unstructured finite element meshes. Given the complexity of real world problems, experience has shown that using commercial mesh generator to create rather simple input geometries is overly complex and slow. As a consequence, significant effort has been put into place to create multiple codes that help assist in the mesh generation and manipulation.more » There are three input means to create a mesh in PROTEUS: UFMESH, GRID, and NEMESH. At present, the UFMESH is a simple way to generate two-dimensional Cartesian and hexagonal fuel assembly geometries. The UFmesh input allows for simple assembly mesh generation while the GRID input allows the generation of Cartesian, hexagonal, and regular triangular structured grid geometry options. The NEMESH is a way for the user to create their own mesh or convert another mesh file format into a PROTEUS input format. Given that one has an input mesh format acceptable for PROTEUS, we have constructed several tools which allow further mesh and geometry construction (i.e. mesh extrusion and merging). This report describes the various mesh tools that are provided with the PROTEUS code giving both descriptions of the input and output. In many cases the examples are provided with a regression test of the mesh tools. The most important mesh tools for any user to consider using are the MT_MeshToMesh.x and the MT_RadialLattice.x codes. The former allows the conversion between most mesh types handled by PROTEUS while the second allows the merging of multiple (assembly) meshes into a radial structured grid. Note that the mesh generation process is recursive in nature and that each input specific for a given mesh tool (such as .axial or .merge) can be used as “mesh” input for any of the mesh tools discussed in this manual.« less

Studies on piston bowl geometries using single blend ratio of various non-edible oils.

PubMed

Viswanathan, Karthickeyan; Pasupathy, Balamurugan

2017-07-01

The depletion of fossil fuels and hike in crude oil prices were some of the main reasons to explore new alternatives from renewable source of energy. This work presents the impact of various bowl geometries on diesel engine with diesel and biodiesel samples. Three non-edible oils were selected, namely pumpkin seed oil, orange oil and neem oil. These oils were converted into respective biodiesel using transesterification process in the presence of catalyst and alcohol. After transesterification process, the oils were termed as pumpkin seed oil methyl ester (PSOME), orange oil methyl ester (OME) and neem oil methyl ester (NOME), respectively. The engine used for experimentation was a single-cylinder four-stroke water-cooled direct-injection diesel engine and loads were applied to the engine using eddy current dynamometer. Two bowl geometries were developed, namely toroidal combustion chamber (TCC) and trapezoidal combustion chamber (TRCC). Also, the engine was inbuilt with hemispherical combustion chamber (HCC). The base line readings were recorded using neat diesel fuel with HCC for various loads. Followed by 20% of biodiesel mixed with 80% neat diesel for all prepared methyl esters and termed as B1 (20% PSOME with 80% diesel), B2 (20% OME with 80% diesel) and B3 (20% NOME with 80% diesel). All fuel samples were tested in HCC, TCC and TRCC bowl geometries under standard injection timing and with compression ratio of 18. Increased brake thermal efficiency and reduced brake specific fuel consumption were observed with diesel in TCC geometry. Also, higher heat release and cylinder pressures with lower ignition delay were recorded with TCC bowl geometry. TCC bowl geometry showed lower CO, HC and smoke emissions with B2 fuel sample than diesel and other biodiesel samples. But, higher NOx emission was observed in HCC and TCC than that in TRCC bowl geometry. Graphical abstract ᅟ.

Study of effects of injector geometry on fuel-air mixing and combustion

NASA Technical Reports Server (NTRS)

Bangert, L. H.; Roach, R. L.

1977-01-01

An implicit finite-difference method has been developed for computing the flow in the near field of a fuel injector as part of a broader study of the effects of fuel injector geometry on fuel-air mixing and combustion. Detailed numerical results have been obtained for cases of laminar and turbulent flow without base injection, corresponding to the supersonic base flow problem. These numerical results indicated that the method is stable and convergent, and that significant savings in computer time can be achieved, compared with explicit methods.

The simulation of the alternate turbopump development high pressure oxygen and fuel turbopumps for the space shuttle main engine using the Shaberth computer program

NASA Technical Reports Server (NTRS)

Mcdonald, Gary H.

1988-01-01

The Space Shuttle Main Engine (SSME) is basically comprised of a combustion chamber and nozzle, high and low pressure oxygen turbopumps and high and low pressure fuel turbopumps. In the current configuration, the high pressure fuel (HPTFP) and high pressure oxygen turbopumps (HPOTP) have experienced a history of ball bearing wear. The wear problem can be attributed to numerous factors including the hydrodynamic axial and radial loads caused by the flow of liquid oxygen and liquid hydrogen through the turbopump impellers and turbine. Also, friction effects between the rolling elements, races, and cage can create thermally induced bearing geometry changes. To alleviate some of the current configuration problems, an alternate turbopump development (ATD) was proposed. However, the ATD HPOTP and HPTFP are constrained to operate interchangeably with the current turbopumps, thus, the operation conditions must be similar. The ATD configuration features a major change in bearings used to support the integrated shaft, impeller, and turbine system. A single ball and single roller will replace the pump-end and turbine and duplex ball bearings. The Shaft-Bearing-Thermal (SHABERTH) computer code was used to model the ATD HPOTP and ATD HPFTP configurations. A two bearing model was used to simulate the HPOTP and HPFTP bearings and shaft geometry. From SHABERTH, a comparison of bearing reaction loads, frictional heat generation rates, and Hertz contact stresses will be attempted with analysis at the 109 percent and 65 percent power levels.

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

NASA Astrophysics Data System (ADS)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-07-01

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D

FUEL ELEMENT SUPPORT

DOEpatents

Wyman, W.L.

1961-06-27

The described cylindrical fuel element has longitudinally spaced sets of short longitudinal ribs circumferentially spaced from one another. The ribs support the fuel element in a coolant tube so that there is an annular space for coolant flow between the fuel element and the interior of the coolant tube. If the fuel element grows as a result of reactor operation, the circumferential distribution of the ribs maintains the uniformity of the annular space between the coolant tube and the fuel element, and the collapsibility of the ribs prevents the fuel element from becoming jammed in the coolant tube.

Thermal breeder fuel enrichment zoning

DOEpatents

Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

1992-01-01

A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Emrich, William J. Jr.

2008-01-21

To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowingmore » hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.« less

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

Low temperature chemical processing of graphite-clad nuclear fuels

DOEpatents

Pierce, Robert A.

2017-10-17

A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

Broad Specification Fuels Combustion Technology Program. Phase 2

DTIC Science & Technology

1990-10-01

4 4C Where: M is the molecular weight of th hxth specie Nt is the mole fraction of the x specie a is the hydrogen to carbon ratio of the fuel...RATIO F’gure 7-15 Idle Emisions Characteristics of Variable Geometry Cornbusuom geometry combustor configurations as well. The remaining performance

77 FR 5418 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-03

... aft fuel system 40 micron fuel filter element with a 10 micron fuel filter element. This proposed AD... fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. This proposed AD would require replacing each forward and aft fuel system 40 micron fuel filter element with a 10 micron fuel filter...

NUCLEAR REACTOR FUEL ELEMENT

DOEpatents

Wheelock, C.W.; Baumeister, E.B.

1961-09-01

A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

Contact behavior modelling and its size effect on proton exchange membrane fuel cell

NASA Astrophysics Data System (ADS)

Qiu, Diankai; Peng, Linfa; Yi, Peiyun; Lai, Xinmin; Janßen, Holger; Lehnert, Werner

2017-10-01

Contact behavior between the gas diffusion layer (GDL) and bipolar plate (BPP) is of significant importance for proton exchange membrane fuel cells. Most current studies on contact behavior utilize experiments and finite element modelling and focus on fuel cells with graphite BPPs, which lead to high costs and huge computational requirements. The objective of this work is to build a more effective analytical method for contact behavior in fuel cells and investigate the size effect resulting from configuration alteration of channel and rib (channel/rib). Firstly, a mathematical description of channel/rib geometry is outlined in accordance with the fabrication of metallic BPP. Based on the interface deformation characteristic and Winkler surface model, contact pressure between BPP and GDL is then calculated to predict contact resistance and GDL porosity as evaluative parameters of contact behavior. Then, experiments on BPP fabrication and contact resistance measurement are conducted to validate the model. The measured results demonstrate an obvious dependence on channel/rib size. Feasibility of the model used in graphite fuel cells is also discussed. Finally, size factor is proposed for evaluating the rule of size effect. Significant increase occurs in contact resistance and porosity for higher size factor, in which channel/rib width decrease.

Low cost, lightweight fuel cell elements

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor)

2001-01-01

New fuel cell elements for use in liquid feed fuel cells are provided. The elements including biplates and endplates are low in cost, light in weight, and allow high efficiency operation. Electrically conductive elements are also a part of the fuel cell elements.

Elucidating bonding preferences in tetrakis(imido)uranate(VI) dianions

NASA Astrophysics Data System (ADS)

Anderson, Nickolas H.; Xie, Jing; Ray, Debmalya; Zeller, Matthias; Gagliardi, Laura; Bart, Suzanne C.

2017-09-01

Actinyl species, [AnO2]2+, are well-known derivatives of the f-block because of their natural occurrence and essential roles in the nuclear fuel cycle. Along with their nitrogen analogues, [An(NR)2]2+, actinyls are characterized by their two strong trans-An-element multiple bonds, a consequence of the inverse trans influence. We report that these robust bonds can be weakened significantly by increasing the number of multiple bonds to uranium, as demonstrated by a family of uranium(VI) dianions bearing four U-N multiple bonds, [M]2[U(NR)4] (M = Li, Na, K, Rb, Cs). Their geometry is dictated by cation coordination and sterics rather than by electronic factors. Multiple bond weakening by the addition of strong π donors has the potential for applications in the processing of high-valent actinyls, commonly found in environmental pollutants and spent nuclear fuels.

PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1981-09-01

This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity.more » The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.« less

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Nuclear reactor composite fuel assembly

DOEpatents

Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

1980-01-01

A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

Method of locating a leaking fuel element in a fast breeder power reactor

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

DOEpatents

Weems, S.J.

1963-09-24

A rib construction for a nuclear-fuel element is described, in which one of three peripherally spaced ribs adjacent to each end of the fuel element is mounted on a radially yielding spring that embraces the fuel element. This spring enables the fuel element to have a good fit with a coolant tube and yet to be easily inserted in and withdrawn from the tube. (AEC)

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

Optimization of polymer electrolyte membrane fuel cell flow channels using a genetic algorithm

NASA Astrophysics Data System (ADS)

Catlin, Glenn; Advani, Suresh G.; Prasad, Ajay K.

The design of the flow channels in PEM fuel cells directly impacts the transport of reactant gases to the electrodes and affects cell performance. This paper presents results from a study to optimize the geometry of the flow channels in a PEM fuel cell. The optimization process implements a genetic algorithm to rapidly converge on the channel geometry that provides the highest net power output from the cell. In addition, this work implements a method for the automatic generation of parameterized channel domains that are evaluated for performance using a commercial computational fluid dynamics package from ANSYS. The software package includes GAMBIT as the solid modeling and meshing software, the solver FLUENT, and a PEMFC Add-on Module capable of modeling the relevant physical and electrochemical mechanisms that describe PEM fuel cell operation. The result of the optimization process is a set of optimal channel geometry values for the single-serpentine channel configuration. The performance of the optimal geometry is contrasted with a sub-optimal one by comparing contour plots of current density, oxygen and hydrogen concentration. In addition, the role of convective bypass in bringing fresh reactant to the catalyst layer is examined in detail. The convergence to the optimal geometry is confirmed by a bracketing study which compares the performance of the best individual to those of its neighbors with adjacent parameter values.

Morphing the feature-based multi-blocks of normative/healthy vertebral geometries to scoliosis vertebral geometries: development of personalized finite element models.

PubMed

Hadagali, Prasannaah; Peters, James R; Balasubramanian, Sriram

2018-03-01

Personalized Finite Element (FE) models and hexahedral elements are preferred for biomechanical investigations. Feature-based multi-block methods are used to develop anatomically accurate personalized FE models with hexahedral mesh. It is tedious to manually construct multi-blocks for large number of geometries on an individual basis to develop personalized FE models. Mesh-morphing method mitigates the aforementioned tediousness in meshing personalized geometries every time, but leads to element warping and loss of geometrical data. Such issues increase in magnitude when normative spine FE model is morphed to scoliosis-affected spinal geometry. The only way to bypass the issue of hex-mesh distortion or loss of geometry as a result of morphing is to rely on manually constructing the multi-blocks for scoliosis-affected spine geometry of each individual, which is time intensive. A method to semi-automate the construction of multi-blocks on the geometry of scoliosis vertebrae from the existing multi-blocks of normative vertebrae is demonstrated in this paper. High-quality hexahedral elements were generated on the scoliosis vertebrae from the morphed multi-blocks of normative vertebrae. Time taken was 3 months to construct the multi-blocks for normative spine and less than a day for scoliosis. Efforts taken to construct multi-blocks on personalized scoliosis spinal geometries are significantly reduced by morphing existing multi-blocks.

Modeling CANDU-6 liquid zone controllers for effects of thorium-based fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

St-Aubin, E.; Marleau, G.

2012-07-01

We use the DRAGON code to model the CANDU-6 liquid zone controllers and evaluate the effects of thorium-based fuels on their incremental cross sections and reactivity worth. We optimize both the numerical quadrature and spatial discretization for 2D cell models in order to provide accurate fuel properties for 3D liquid zone controller supercell models. We propose a low computer cost parameterized pseudo-exact 3D cluster geometries modeling approach that avoids tracking issues on small external surfaces. This methodology provides consistent incremental cross sections and reactivity worths when the thickness of the buffer region is reduced. When compared with an approximate annularmore » geometry representation of the fuel and coolant region, we observe that the cluster description of fuel bundles in the supercell models does not increase considerably the precision of the results while increasing substantially the CPU time. In addition, this comparison shows that it is imperative to finely describe the liquid zone controller geometry since it has a strong impact of the incremental cross sections. This paper also shows that liquid zone controller reactivity worth is greatly decreased in presence of thorium-based fuels compared to the reference natural uranium fuel, since the fission and the fast to thermal scattering incremental cross sections are higher for the new fuels. (authors)« less

On the effects of grid ill-conditioning in three dimensional finite element vector potential magnetostatic field computations

NASA Technical Reports Server (NTRS)

Wang, R.; Demerdash, N. A.

1990-01-01

The effects of finite element grid geometries and associated ill-conditioning were studied in single medium and multi-media (air-iron) three dimensional magnetostatic field computation problems. The sensitivities of these 3D field computations to finite element grid geometries were investigated. It was found that in single medium applications the unconstrained magnetic vector potential curl-curl formulation in conjunction with first order finite elements produce global results which are almost totally insensitive to grid geometries. However, it was found that in multi-media (air-iron) applications first order finite element results are sensitive to grid geometries and consequent elemental shape ill-conditioning. These sensitivities were almost totally eliminated by means of the use of second order finite elements in the field computation algorithms. Practical examples are given in this paper to demonstrate these aspects mentioned above.

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. SHOWS AIR FORCE MAN AT EDGE OF TANK. INEL PHOTO NUMBER 65-6176, TAKEN NOVEMBER 10, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanson, A. L.; Diamond, D.

2013-10-31

The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposedmore » LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.« less

A Brief History of Non-Euclidean Geometry

ERIC Educational Resources Information Center

Marshall, Daniel; Scott, Paul

2004-01-01

Around 300 BC, Euclid wrote "The Elements", a major treatise on the geometry of the time, and what would be considered "geometry" for many years after. Arguably "The Elements" is the second most read book of the western world, falling short only to The Bible. In his book, Euclid states five postulates of geometry which he uses as the foundation…

NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Nuclear reactor

DOEpatents

Thomson, Wallace B.

2004-03-16

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

Optimisation of the geometry of the drill bit and process parameters for cutting hybrid composite/metal structures in new aircrafts

NASA Astrophysics Data System (ADS)

Isbilir, Ozden

Owing to their desirable strength-to-weight characteristics, carbon fibre reinforced polymer composites have been favourite materials for structural applications in different industries such as aerospace, transport, sports and energy. They provide a weight reduction in whole structure and consequently decrease fuel consumption. The use of lightweight materials such as titanium and its alloys in modern aircrafts has also increased significantly in the last couple of decades. Titanium and its alloys offer high strength/weight ratio, high compressive and tensile strength at high temperatures, low density, excellent corrosion resistance, exceptional erosion resistance, superior fatigue resistance and relatively low modulus of elasticity. Although composite/metal hybrid structures are increasingly used in airframes nowadays, number of studies regarding drilling of composite/metal stacks is very limited. During drilling of multilayer materials different problems may arise due to very different attributes of these materials. Machining conditions of drilling such structures play an important role on tool wear, quality of holes and cost of machining.. The research work in this thesis is aimed to investigate drilling of CFRP/Ti6Al4V hybrid structure and to optimize process parameters and drill geometry. The research work consist complete experimental study including drilling tests, in-situ and post measurements and related analysis; and finite element analysis including fully 3-D finite element models. The experimental investigations focused on drilling outputs such as thrust force, torque, delamination, burr formation, surface roughness and tool wear. An algorithm was developed to analyse drilling induced delamination quantitatively based on the images. In the numerical analysis, novel 3-D finite element models of drilling of CFRP, Ti6Al4V and CFRP/Ti6Al4V hybrid structure were developed with the use of 3-D complex drill geometries. A user defined subroutine was developed to model material and failure behaviour of CFRP. The effects of process parameters on drilling outputs have been investigated and compared with the experimental results. The influences of drill bit geometries have been simulated in this study..

Components for digitally controlled aircraft engines

NASA Technical Reports Server (NTRS)

Meador, J. D.

1981-01-01

Control system components suitable for use in digital electronic control systems are defined. Compressor geometry actuation concepts and fuel handling system concepts suitable for use in large high performance turbofan/turbojet engines are included. Eight conceptual system designs were formulated for the actuation of the compressor geometry. Six conceptual system designs were formulated for the engine fuel handling system. Assessment criteria and weighting factors were established and trade studies performed on their candidate systems to establish the relative merits of the various concepts. Fuel pumping and metering systems for small turboshaft engines were also studied. Seven conceptual designs were formulated, and trade studies performed. A simplified bypassing fuel metering scheme was selected and a preliminary design defined.

Neutronic fuel element fabrication

DOEpatents

Korton, George

2004-02-24

This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

Nuclear reactor control

DOEpatents

Cawley, William E.; Warnick, Robert F.

1982-01-01

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Neutron Tomography Using Mobile Neutron Generators for Assessment of Void Distributions in Thermal Hydraulic Test Loops

NASA Astrophysics Data System (ADS)

Andersson, P.; Bjelkenstedt, T.; Sundén, E. Andersson; Sjöstrand, H.; Jacobsson-Svärd, S.

Detailed knowledge of the lateral distribution of steam (void) and water in a nuclear fuel assembly is of great value for nuclear reactor operators and fuel manufacturers, with consequences for both reactor safety and economy of operation. Therefore, nuclear relevant two-phase flows are being studied at dedicated thermal-hydraulic test loop, using two-phase flow systems ranging from simplified geometries such as heated circular pipes to full scale mock-ups of nuclear fuel assemblies. Neutron tomography (NT) has been suggested for assessment of the lateral distribution of steam and water in such test loops, motivated by a good ability of neutrons to penetrate the metallic structures of metal pipes and nuclear fuel rod mock-ups, as compared to e.g. conventional X-rays, while the liquid water simultaneously gives comparatively good contrast. However, these stationary test loops require the measurement setup to be mobile, which is often not the case for NT setups. Here, it is acknowledged that fast neutrons of 14 MeV from mobile neutron generators constitute a viable option for a mobile NT system. We present details of the development of neutron tomography for this purpose at the division of Applied Nuclear Physics at Uppsala University. Our concept contains a portable neutron generator, exploiting the fusion reaction of deuterium and tritium, and a detector with plastic scintillator elements designed to achieveadequate spatial and energy resolution, all mounted in a light-weight frame without collimators or bulky moderation to allow for a mobile instrument that can be moved about the stationary thermal hydraulic test sections. The detector system stores event-to-event pulse-height information to allow for discrimination based on the energy deposition in the scintillator elements.

77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

Fuel pumping system and method

DOEpatents

Shafer, Scott F [Morton, IL; Wang, Lifeng ,

2006-12-19

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

Fuel Pumping System And Method

DOEpatents

Shafer, Scott F.; Wang, Lifeng

2005-12-13

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

A comparison of FE beam and continuum elements for typical nitinol stent geometries

NASA Astrophysics Data System (ADS)

Ballew, Wesley; Seelecke, Stefan

2009-03-01

With interest in improved efficiency and a more complete description of the SMA material, this paper compares finite element (FE) simulations of typical stent geometries using two different constitutive models and two different element types. Typically, continuum elements are used for the simulation of stents, for example the commercial FE software ANSYS offers a continuum element based on Auricchio's SMA model. Almost every stent geometry, however, is made up of long and slender components and can be modeled more efficiently, in the computational sense, with beam elements. Using the ANSYS user programmable material feature, we implement the free energy based SMA model developed by Mueller and Seelecke into the ANSYS beam element 188. Convergence behavior for both, beam and continuum formulations, is studied in terms of element and layer number, respectively. This is systematically illustrated first for the case of a straight cantilever beam under end loading, and subsequently for a section of a z-bend wire, a typical stent sub-geometry. It is shown that the computation times for the beam element are reduced to only one third of those of the continuum element, while both formulations display a comparable force/displacement response.

Means for supporting fuel elements in a nuclear reactor

DOEpatents

Andrews, Harry N.; Keller, Herbert W.

1980-01-01

A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

Fuel assembly for nuclear reactors

DOEpatents

Creagan, Robert J.; Frisch, Erling

1977-01-01

A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

An Experimental Investigation of Self-Excited Combustion Dynamics in a Single Element Lean Direct Injection (LDI) Combustor

NASA Astrophysics Data System (ADS)

Gejji, Rohan M.

The management of combustion dynamics in gas turbine combustors has become more challenging as strict NOx/CO emission standards have led to engine operation in a narrow, lean regime. While premixed or partially premixed combustor configurations such as the Lean Premixed Pre-vaporized (LPP), Rich Quench Lean burn (RQL), and Lean Direct Injection (LDI) have shown a potential for reduced NOx emissions, they promote a coupling between acoustics, hydrodynamics and combustion that can lead to combustion instabilities. These couplings can be quite complex, and their detailed understanding is a pre-requisite to any engine development program and for the development of predictive capability for combustion instabilities through high-fidelity models. The overarching goal of this project is to assess the capability of high-fidelity simulation to predict combustion dynamics in low-emissions gas turbine combustors. A prototypical lean-direct-inject combustor was designed in a modular configuration so that a suitable geometry could be found by test. The combustor comprised a variable length air plenum and combustion chamber, air swirler, and fuel nozzle located inside a subsonic venturi. The venturi cross section and the fuel nozzle were consistent with previous studies. Test pressure was 1 MPa and variables included geometry and acoustic resonance, inlet temperatures, equivalence ratio, and type of liquid fuel. High-frequency pressure measurements in a well-instrumented metal chamber yielded frequencies and mode shapes as a function of inlet air temperature, equivalence ratio, fuel nozzle placement, and combustor acoustic resonances. The parametric survey was a significant effort, with over 105 tests on eight geometric configurations. A good dataset was obtained that could be used for both operating-point-dependent quantitative comparisons, and testing the ability of the simulation to predict more global trends. Results showed a very strong dependence of instability amplitude on the geometric configuration of the combustor, i.e., its acoustic resonance characteristics, with measured pressure fluctuation amplitudes ranged from 5 kPa (0.5% of mean pressure) to 200 kPa ( 20% of mean pressure) depending on combustor geometry. The stability behavior also showed a consistent and pronounced dependence on equivalence ratio and inlet air temperature. Instability amplitude increased with higher equivalence ratio and with lower inlet air temperature. A pronounced effect of fuel nozzle location on the combustion dynamics was also observed. Combustion instabilities with the fuel nozzle at the throat of the venturi throat were stronger than in the configuration with fuel nozzle 2.6 mm upstream of the nozzle. A second set of dynamics data was based on high-response-rate laser-based combustion diagnostics using an optically accessible combustor section. High-frequency measurements of OH*-chemiluminescence and OH-PLIF and velocity fields using PIV were obtained at a relatively stable, low equivalence ratio case and a less stable case at higher equivalence ratio. PIV measurements were performed at 5 kHz for non-reacting flow but glare from the cylindrical quartz chamber limited the field of view to a small region in the combustor. Quantitative and qualitative comparisons were made for five different combinations of geometry and operating condition that yielded discriminating stability behavior in the experiment with simulations that were carried out concurrently. Comparisons were made on the basis of trends and pressure mode data as well as with OH-PLIF measurements for the baseline geometry at equivalence ratios of 0.44 and 0.6. Overall, the ability of the simulation to match experimental data and trends was encouraging. Dynamic Mode Decomposition (DMD) analysis was performed on two sets of computations - a global 2-step chemistry mechanism and an 18-step chemistry mechanism - and the OH-PLIF images to allow comparison of dynamic patterns of heat release and OH distribution in the combustion zone. The DMD analysis was able to identify similar dominant unstable modes in the combustor. Recommendations for future work are based on the continued requirement for quantitative and spatio-temporally resolved data for direct comparison with computational efforts to develop predictive capabilities for combustion instabilities at relevant operating conditions. Discriminating instability behavior for the prototypical combustor demonstrated in this study is critical for any robust validation effort Unit physics based scaling of the current effort to multi-element combustors along with improvement in diagnostic techniques and analysis efforts are recommended for advancement in understanding of the complex physics in the multi-phase, three dimensional and turbulent combustion processes in the LDI combustor.

Current status of the development of high density LEU fuel for Russian research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

78 FR 17591 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-22

... aft fuel system 40 micron fuel filter element with a 10 micron nominal (40 micron absolute) fuel filter element. This AD was prompted by a National Transportation Safety Board (NTSB) review of in... helicopters with a fuel system 40 micron fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. That...

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

NASA Astrophysics Data System (ADS)

Mella, R.; Wenman, M. R.

2013-06-01

Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of different FE and finite difference models. Non-linear mechanical behaviour of the fuel and cladding including, fuel creep and swelling and cladding creep and plasticity each with dependencies on a variety of different properties. A fission gas release model which takes inputs from first principles calculations. Explicitly integrated inventory calculations performed in a coupled manner. Freedom to model steady state and transient behaviour using implicit time integration. The whole pin geometry is considered over an entire typical fuel life. The model showed by examination of normal operation and a subsequent transient chosen for software demonstration purposes: ABAQUS may be a sufficiently flexible platform to develop a complete and verified fuel performance code. The importance and effectiveness of the geometry of the fuel spacer pellets was characterised. The fuels performance under normal conditions (high friction no power spikes) would not suggest serious degradation of the cladding in fuel life. Large plastic strains were found when pellet bonding was strong, these would appear at all pellets cladding triple points and all pellet radial crack and cladding interfaces thus showing a possible axial direction to cracks forming from ductility exhaustion.

Automotive gas turbine fuel control

NASA Technical Reports Server (NTRS)

Gold, H. (Inventor)

1978-01-01

A fuel control system is reported for automotive-type gas turbines and particulary advanced gas turbines utilizing variable geometry components to improve mileage and reduce pollution emission. The fuel control system compensates for fuel density variations, inlet temperature variations, turbine vane actuation, acceleration, and turbine braking. These parameters are utilized to control various orifices, spool valves and pistons.

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

DOEpatents

Szilard, L.; Young, G.J.

1958-03-01

This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

Leaf-architectured 3D Hierarchical Artificial Photosynthetic System of Perovskite Titanates Towards CO2 Photoreduction Into Hydrocarbon Fuels

PubMed Central

Zhou, Han; Guo, Jianjun; Li, Peng; Fan, Tongxiang; Zhang, Di; Ye, Jinhua

2013-01-01

The development of an “artificial photosynthetic system” (APS) having both the analogous important structural elements and reaction features of photosynthesis to achieve solar-driven water splitting and CO2 reduction is highly challenging. Here, we demonstrate a design strategy for a promising 3D APS architecture as an efficient mass flow/light harvesting network relying on the morphological replacement of a concept prototype-leaf's 3D architecture into perovskite titanates for CO2 photoreduction into hydrocarbon fuels (CO and CH4). The process uses artificial sunlight as the energy source, water as an electron donor and CO2 as the carbon source, mimicking what real leaves do. To our knowledge this is the first example utilizing biological systems as “architecture-directing agents” for APS towards CO2 photoreduction, which hints at a more general principle for APS architectures with a great variety of optimized biological geometries. This research would have great significance for the potential realization of global carbon neutral cycle. PMID:23588925

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montierth, Leland M.

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

Validating the BISON fuel performance code to integral LWR experiments

DOE PAGES

Williamson, R. L.; Gamble, K. A.; Perez, D. M.; ...

2016-03-24

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to datemore » for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Study of Geometric Stability and Structural Integrity of Self-Healing Glass Seal System Used in Solid Oxide Fuel Cells

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Wenning N.; Sun, Xin; Khaleel, Mohammad A.

A self-healing glass seal has the potential of restoring its mechanical properties upon being reheated to SOFC stack operating temperature, even when it has experienced some cooling induced damage/cracking at room temperature. Such a self-healing feature is desirable for achieving high seal reliability during thermal cycling. On the other hand, self-healing glass is also characterized by its low mechanical stiffness and high creep rate at the typical operating temperature of SOFCs. Therefore, geometry stability and structural integrity of the glass seal system becomes critical to its successful application in SOFCs. In this paper, the geometry stability of the self-healing glassmore » and the influence of various interfacial conditions of ceramic stoppers with the PEN, IC, and glass seal on the structural integrity of the glass seal during the operating and cooling down processes are studied using finite element analyses. For this purpose, the test cell used in the leakage tests for compliant glass seals conducted at PNNL is taken as the initial modeling geometry. The effect of the ceramic stopper on the geometry stability of the self-healing glass sealants is studied first. Two interfacial conditions of the ceramic stopper and glass seals, i.e., bonded (strong) or un-bonded (weak), are considered. Then the influences of interfacial strengths at various interfaces, i.e., stopper/glass, stopper/PEN, as well as stopper/IC plate, on the geometry stability and reliability of glass during the operating and cooling processes are examined.« less

Evaluation of the finite element fuel rod analysis code (FRANCO)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, K.; Feltus, M.A.

1994-12-31

Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-11-07

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-01-01

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

DOEpatents

Stengel, F.G.

1963-12-24

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

Direct hydrocarbon fuel cells

DOEpatents

Barnett, Scott A.; Lai, Tammy; Liu, Jiang

2010-05-04

The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

DOEpatents

Wigner, E.P.; Young, G.J.; Weinberg, A.M.

1961-06-27

A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-16

A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

SLUG HANDLING DEVICES

DOEpatents

Gentry, J.R.

1958-09-16

A device is described for handling fuel elements of a neutronic reactor. The device consists of two concentric telescoped contalners that may fit about the fuel element. A number of ratchet members, equally spaced about the entrance to the containers, are pivoted on the inner container and spring biased to the outer container so thnt they are forced to hear against and hold the fuel element, the weight of which tends to force the ratchets tighter against the fuel element. The ratchets are released from their hold by raising the inner container relative to the outer memeber. This device reduces the radiation hazard to the personnel handling the fuel elements.

FUEL ELEMENTS FOR NUCLEAR REACTORS

DOEpatents

Blainey, A.; Lloyd, H.

1961-07-11

A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trammell, Michael P; Jolly, Brian C; Miller, James Henry

ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

Fuel element concept for long life high power nuclear reactors

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ELEMENT HOLDER, TRIP MECHANISM COVER, AND OTHER DETAILS. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-3. INEL INDEX CODE NUMBER: 075 0701 60 851 151977. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Possible consequences of operation with KIVN fuel elements in K Zircaloy process tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlson, P.A.

1963-08-06

From considerations of the results of experimental simulations of non-axial placement of fuel elements in process tubes and in-reactor experience, it is concluded that the ultimate outcome of a charging error which results in operation with one or more unsupported fuel elements in a K Zircaloy-2 process tube would be multiple fuel failure and failure of the process tube. The outcome of the accident is determined by the speed with which the fuel failure is detected and the reactor is shut down. The release of fission products would be expected to be no greater than that which has occurred followingmore » severe fuel failure incidents. The highest probability for fission product release occurs during the discharge of failed fuel elements, when a small fraction of the exposed uranium of the fuel element may be oxidized when exposed to air before the element falls into the water-filled discharge chute. The confinement and fog spray facilities were installed to reduce the amount of fission products which might escape from the reactor building after such an event.« less

Scramjet fuel injector design parameters and considerations: Development of a two-dimensional tangential fuel injector with constant pressure at the flame

NASA Technical Reports Server (NTRS)

Agnone, A. M.

1972-01-01

The factors affecting a tangential fuel injector design for scramjet operation are reviewed and their effect on the efficiency of the supersonic combustion process is evaluated using both experimental data and theoretical predictions. A description of the physical problem of supersonic combustion and method of analysis is followed by a presentation and evaluation of some standard and exotic types of fuel injectors. Engineering fuel injector design criteria and hydrogen ignition schemes are presented along with a cursory review of available experimental data. A two-dimensional tangential fuel injector design is developed using analyses as a guide in evaluating the effects on the combustion process of various initial and boundary conditions including splitter plate thickness, injector wall temperature, pressure gradients, etc. The fuel injector wall geometry is shaped so as to maintain approximately constant pressure at the flame as required by a cycle analysis. A viscous characteristics program which accounts for lateral as well as axial pressure variations due to the mixing and combustion process is used in determining the wall geometry.

Rack for storing spent nuclear fuel elements

DOEpatents

Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

1978-06-20

A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

A Jacobian-free Newton Krylov method for mortar-discretized thermomechanical contact problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hansen, Glen, E-mail: Glen.Hansen@inl.gov

2011-07-20

Multibody contact problems are common within the field of multiphysics simulation. Applications involving thermomechanical contact scenarios are also quite prevalent. Such problems can be challenging to solve due to the likelihood of thermal expansion affecting contact geometry which, in turn, can change the thermal behavior of the components being analyzed. This paper explores a simple model of a light water reactor nuclear fuel rod, which consists of cylindrical pellets of uranium dioxide (UO{sub 2}) fuel sealed within a Zircalloy cladding tube. The tube is initially filled with helium gas, which fills the gap between the pellets and cladding tube. Themore » accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel performance, including cladding stress and behavior under irradiated conditions, which are factors that affect the lifetime of the fuel. The thermomechanical contact approach developed here is based on the mortar finite element method, where Lagrange multipliers are used to enforce weak continuity constraints at participating interfaces. In this formulation, the heat equation couples to linear mechanics through a thermal expansion term. Lagrange multipliers are used to formulate the continuity constraints for both heat flux and interface traction at contact interfaces. The resulting system of nonlinear algebraic equations are cast in residual form for solution of the transient problem. A Jacobian-free Newton Krylov method is used to provide for fully-coupled solution of the coupled thermal contact and heat equations.« less

A Jacobian-Free Newton Krylov Method for Mortar-Discretized Thermomechanical Contact Problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glen Hansen

2011-07-01

Multibody contact problems are common within the field of multiphysics simulation. Applications involving thermomechanical contact scenarios are also quite prevalent. Such problems can be challenging to solve due to the likelihood of thermal expansion affecting contact geometry which, in turn, can change the thermal behavior of the components being analyzed. This paper explores a simple model of a light water reactor nuclear reactor fuel rod, which consists of cylindrical pellets of uranium dioxide (UO2) fuel sealed within a Zircalloy cladding tube. The tube is initially filled with helium gas, which fills the gap between the pellets and cladding tube. Themore » accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel performance, including cladding stress and behavior under irradiated conditions, which are factors that affect the lifetime of the fuel. The thermomechanical contact approach developed here is based on the mortar finite element method, where Lagrange multipliers are used to enforce weak continuity constraints at participating interfaces. In this formulation, the heat equation couples to linear mechanics through a thermal expansion term. Lagrange multipliers are used to formulate the continuity constraints for both heat flux and interface traction at contact interfaces. The resulting system of nonlinear algebraic equations are cast in residual form for solution of the transient problem. A Jacobian-free Newton Krylov method is used to provide for fully-coupled solution of the coupled thermal contact and heat equations.« less

Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

DOEpatents

Sease, J.D.; Harrington, F.E.

1973-12-11

Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

NASA Technical Reports Server (NTRS)

Mireles, O. R.; Broadway, J.; Hickman, R.

2014-01-01

Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

Preparation of high temperature gas-cooled reactor fuel element

DOEpatents

Bradley, Ronnie A.; Sease, John D.

1976-01-01

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

Nuclear fuel element

DOEpatents

Zocher, Roy W.

1991-01-01

A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

Operating condition and geometry effects on low-frequency afterburner combustion instability in a turbofan at altitude

NASA Technical Reports Server (NTRS)

Cullom, R. R.; Johnsen, R. L.

1979-01-01

Three afterburner configurations were tested in a low-bypass-ratio turbofan engine to determine the effect of various fuel distributions, inlet conditions, flameholder geometry, and fuel injection location on combustion instability. Tests were conducted at simulated flight conditions of Mach 0.75 and 1.3 at altitudes from 11,580 to 14,020 m (38,000 to 46,000 ft). In these tests combustion instability with frequency from 28 to 90 Hz and peak-to-peak pressure amplitude up to 46.5 percent of the afterburner inlet total pressure level was encountered. Combustion instability was suppressed in these tests by varying the fuel distribution in the afterburner.

Dumbo: A pachydermal rocket motor

NASA Technical Reports Server (NTRS)

Kirk, Bill

1991-01-01

A brief historical account is given of the Dumbo nuclear reactor, a type of folded flow reactor that could be used for rocket propulsion. Much of the information is given in viewgraph form. Viewgraphs show details of the reactor system, fuel geometry, and key characteristics of the system (folded flow, use of fuel washers, large flow area, small fuel volume, hybrid modulator, and cermet fuel).

Construction of hexahedral finite element mesh capturing realistic geometries of a petroleum reserve

DOE PAGES

Park, Byoung Yoon; Roberts, Barry L.; Sobolik, Steven R.

2017-07-27

The three-dimensional finite element mesh capturing realistic geometries of the Bayou Choctaw site has been constructed using the sonar and seismic survey data obtained from the field. The mesh consists of hexahedral elements because the salt constitutive model is coded using hexahedral elements. Various ideas and techniques to construct finite element mesh capturing artificially and naturally formed geometries are provided. The techniques to reduce the number of elements as much as possible to save on computer run time while maintaining the computational accuracy is also introduced. The steps and methodologies could be applied to construct the meshes of Big Hill,more » Bryan Mound, and West Hackberry strategic petroleum reserve sites. The methodology could be applied to the complicated shape masses for various civil and geological structures.« less

Construction of hexahedral finite element mesh capturing realistic geometries of a petroleum reserve

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, Byoung Yoon; Roberts, Barry L.; Sobolik, Steven R.

The three-dimensional finite element mesh capturing realistic geometries of the Bayou Choctaw site has been constructed using the sonar and seismic survey data obtained from the field. The mesh consists of hexahedral elements because the salt constitutive model is coded using hexahedral elements. Various ideas and techniques to construct finite element mesh capturing artificially and naturally formed geometries are provided. The techniques to reduce the number of elements as much as possible to save on computer run time while maintaining the computational accuracy is also introduced. The steps and methodologies could be applied to construct the meshes of Big Hill,more » Bryan Mound, and West Hackberry strategic petroleum reserve sites. The methodology could be applied to the complicated shape masses for various civil and geological structures.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Konyashov, Vadim V.; Krasnov, Alexander M.

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. Anmore » approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)« less

System for conversion between the boundary representation model and a constructive solid geometry model of an object

DOEpatents

Christensen, Noel C.; Emery, James D.; Smith, Maurice L.

1988-04-05

A system converts from the boundary representation of an object to the constructive solid geometry representation thereof. The system converts the boundary representation of the object into elemental atomic geometrical units or I-bodies which are in the shape of stock primitives or regularized intersections of stock primitives. These elemental atomic geometrical units are then represented in symbolic form. The symbolic representations of the elemental atomic geometrical units are then assembled heuristically to form a constructive solid geometry representation of the object usable for manufacturing thereof. Artificial intelligence is used to determine the best constructive solid geometry representation from the boundary representation of the object. Heuristic criteria are adapted to the manufacturing environment for which the device is to be utilized. The surface finish, tolerance, and other information associated with each surface of the boundary representation of the object are mapped onto the constructive solid geometry representation of the object to produce an enhanced solid geometry representation, particularly useful for computer-aided manufacture of the object.

Benchmark Wall Heat Flux Data for a GO2/GH2 Single Element Combustor

NASA Technical Reports Server (NTRS)

Marshall, William M.; Pal, Sibtosh; Woodward, Roger d.; Santoro, Robert J.

2005-01-01

Wall heat flux measurements in a 1.5 in. diameter circular cross-section rocket chamber for a uni-element shear coaxial injector element operating on gaseous oxygen (GOz)/gaseous hydrogen (GH,) propellants are presented. The wall heat flux measurements were made using arrays of Gardon type heat flux gauges and coaxial thermocouple instrumentation. Wall heat flux measurements were made for two cases. For the first case, GOZ/GHz oxidizer-rich (O/F=l65) and fuel-rich preburners (O/F=1.09) integrated with the main chamber were utilized to provide vitiated hot fuel and oxidizer to the study shear coaxial injector element. For the second case, the preburners were removed and ambient temperature gaseous oxygen/gaseous hydrogen propellants were supplied to the study injector. Experiments were conducted at four chamber pressures of 750, 600, 450 and 300psia for each case. The overall mixture ratio for the preburner case was 6.6, whereas for the ambient propellant case, the mixture ratio was 6.0. Total propellant flow was nominally 0.27-0.29 Ibm/s for the 750 psia case with flowrates scaled down linearly for lower chamber pressures. The axial heat flux profile results for both the preburner and ambient propellant cases show peak heat flux levels a t axial locations between 2.0 and 3.0 in. from the injector face. The maximum heat flux level was about two times greater for the preburner case. This is attributed to the higher injector fuel-to-oxidizer momentum flux ratio that promotes mixing and higher initial propellant temperature for the preburner case which results in a shorter reaction zone. The axial heat flux profiles were also scaled with respect to the chamber pressure to the power 0.8. The results at the four chamber pressures for both cases collapsed to a single profile indicating that at least to first approximation, the basic fluid dynamic structures in the flow field are pressure independent as long as the chamber/njector/nozzle geometry and injection velocities remain the same.

Photographic combustion characterization of LOX/Hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1980-01-01

One hundred twenty-seven tests were conducted over a chamber pressure range of 125-1500 psia, a fuel temperature range of -245 F to 158 F, and a fuel velocity range of 48-707 ft/sec to demonstrate the advantages and limitations of using high speed photography to identify potential combustion anomalies such as pops, fuel freezing, reactive stream separation and carbon formations. Combustion evaluation criteria were developed to guide selection of the fuels, injector elements, and operating conditions for testing. Separate criteria were developed for fuel and injector element selection and evaluation. The photographic test results indicated conclusively that injector element type and design directly influence carbon formation. Unlike spray fan, impingement elements reduce carbon formation because they induce a relatively rapid near zone fuel vaporization rate. Coherent jet impingement elements, on the other hand, exhibit increased carbon formation.

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

Debonding Stress Concentrations in a Pressurized Lobed Sandwich-Walled Generic Cryogenic Tank

NASA Technical Reports Server (NTRS)

Ko, William L.

2004-01-01

A finite-element stress analysis has been conducted on a lobed composite sandwich tank subjected to internal pressure and cryogenic cooling. The lobed geometry consists of two obtuse circular walls joined together with a common flat wall. Under internal pressure and cryogenic cooling, this type of lobed tank wall will experience open-mode (a process in which the honeycomb is stretched in the depth direction) and shear stress concentrations at the junctures where curved wall changes into flat wall (known as a curve-flat juncture). Open-mode and shear stress concentrations occur in the honeycomb core at the curve-flat junctures and could cause debonding failure. The levels of contributions from internal pressure and temperature loading to the open-mode and shear debonding failure are compared. The lobed fuel tank with honeycomb sandwich walls has been found to be a structurally unsound geometry because of very low debonding failure strengths. The debonding failure problem could be eliminated if the honeycomb core at the curve-flat juncture is replaced with a solid core.

Characterization of Swirl-Venturi Lean Direct Injection Designs for Aviation Gas-Turbine Combustion

NASA Technical Reports Server (NTRS)

Heath, Christopher M.

2013-01-01

Injector geometry, physical mixing, chemical processes, and engine cycle conditions together govern performance, operability and emission characteristics of aviation gas-turbine combustion systems. The present investigation explores swirl-venturi lean direct injection combustor fundamentals, characterizing the influence of key geometric injector parameters on reacting flow physics and emission production trends. In this computational study, a design space exploration was performed using a parameterized swirl-venturi lean direct injector model. From the parametric geometry, 20 three-element lean direct injection combustor sectors were produced and simulated using steady-state, Reynolds-averaged Navier-Stokes reacting computations. Species concentrations were solved directly using a reduced 18-step reaction mechanism for Jet-A. Turbulence closure was obtained using a nonlinear ?-e model. Results demonstrate sensitivities of the geometric perturbations on axially averaged flow field responses. Output variables include axial velocity, turbulent kinetic energy, static temperature, fuel patternation and minor species mass fractions. Significant trends have been reduced to surrogate model approximations, intended to guide future injector design trade studies and advance aviation gas-turbine combustion research.

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

TECHNICAL SCOPE OF GAS-COOLED REACTOR FUEL ELEMENT IRRADIATION PROGRAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

A set of 55 experiments hss been outiined to provide a minimum irradiation program for selection of UO/sub 2/, pellet geometry and fabricntion techniques, and canning technology. These experiments fall into three catagories: prototype: untts in which radial dimension and heat fluxes sre close to proposed design values, but irradiation times are long; reduced-size prototype for accelerated tests in which most variables will be studied; and miniaurized pellet irradiation to obtain high burnup for fission gas release studies. Reactor space has been found generally available and several installations are now examining their capabilities to participate in the program. A tentativemore » schedule has been drawn to illustrate the feasibility of the program. (auth)« less

Fuel handling apparatus for a nuclear reactor

DOEpatents

Hawke, Basil C.

1987-01-01

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Drying results of K-Basin fuel element 1990 (Run 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, S.C.; Abrefah, J.; Klinger, G.S.

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtainedmore » from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.« less

FUEL ELEMENT FOR NEUTRONIC REACTORS

DOEpatents

Evans, T.C.; Beasley, E.G.

1961-01-17

A fuel element for neutronic reactors, particularly the gas-cooled type of reactor, is described. The element comprises a fuel-bearing plate rolled to form a cylinder having a spiral passageway passing from its periphery to its center. In operation a coolant is admitted to the passageway at the periphery of the element, is passed through the spiral passageway, and emerges into a central channel defined by the inner turn of the rolled plate. The advantage of the element is that the fully heated coolant (i.e., coolant emerging into the central channel) is separated and thus insulated from the periphery of the element, which may be in contact with a low-temperature moderator, by the intermediate turns of the spiral fuel element.

Beam heated linear theta-pinch device for producing hot plasmas

DOEpatents

Bohachevsky, Ihor O.

1981-01-01

A device for producing hot plasmas comprising a single turn theta-pinch coil, a fast discharge capacitor bank connected to the coil, a fuel element disposed along the center axis of the coil, a predetermined gas disposed within the theta-pinch coil, and a high power photon, electron or ion beam generator concentrically aligned to the theta-pinch coil. Discharge of the capacitor bank generates a cylindrical plasma sheath within the theta-pinch coil which heats the outer layer of the fuel element to form a fuel element plasma layer. The beam deposits energy in either the cylindrical plasma sheath or the fuel element plasma layer to assist the implosion of the fuel element to produce a hot plasma.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

Evaluation of a staged fuel combustor for turboprop engines

NASA Technical Reports Server (NTRS)

Verdouw, A. J.

1976-01-01

Proposed EPA emission regulations require emission reduction by 1979 for various gas turbine engine classes. Extensive combustion technology advancements are required to meet the proposed regulations. The T56 turboprop engine requires CO, UHC, and smoke reduction. A staged fuel combustor design was tested on a combustion rig to evaluate emission reduction potential in turboprop engines from fuel zoning. The can-type combustor has separately fueled-pilot and main combustion zones in series. The main zone fueling system was arranged for potential incorporation into the T56 with minor or no modifications to the basic engine. Three combustor variable geometry systems were incorporated to evaluate various airflow distributions. Emission results with fixed geometry operation met all proposed EPA regulations over the EPA LTO cycle. CO reduction was 82 percent, UHC reduction was 96 percent, and smoke reduction was 84 percent. NOx increased 14 percent over the LTO cycle. At high power, NOx reduction was 40 to 55 percent. This NOx reduction has potential application to stationary gas turbine powerplants which have different EPA regulations.

PUSH-PULL POWER REACTOR

DOEpatents

Froman, D.K.

1959-02-24

Power generating nuclear reactors of the homogeneous liquid fuel type are discussed. The apparatus utilizes two identical reactors interconnected by conduits through heat exchanging apparatus. Each reactor contains a critical geometry region and a vapor region separated from the critical region by a baffle. When the liquid in the first critical region becomes critical, the vapor pressure above the fuel is increased due to the rise in the temperature until it forces the liquid fuel out of the first critical region through the heat exchanger and into the second critical region, which is at a lower temperature and consequently a lower vapor pressure. The above reaction is repeated in the second critical region and the liquid fuel is forced back into the first critical region. In this manner criticality is achieved alternately in each critical region and power is extracted by the heat exchanger from the liquid fuel passing therethrough. The vapor region and the heat exchanger have a non-critical geometry and reactivity control is effected by conventional control rods in the critical regions.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

Effects of Passive Fuel-Air Mixing Control on Burner Emissions Via Lobed Fuel Injectors

NASA Technical Reports Server (NTRS)

Mitchell, M. G.; Smith, O. I.; Karagozian, A. R.

1999-01-01

The present experimental study examines the effects of differing levels of passive fuel-air premixing on flame structures and their associated NO(x) and CO emissions. Four alternative fuel injector geometries were explored, three of which have lobed shapes. These lobed injectors mix fuel and air and strain species inter-faces to differing extents due to streamwise vorticity generation, thus creating different local or core equivalence ratios within flow regions upstream of flame ignition and stabilization. Prior experimental studies of two of these lobed injector flowfields focused on non-reactive mixing characteristics and emissions measurements for the case where air speeds were matched above and below the fuel injector, effectively generating stronger streamwise vorticity than spanwise vorticity. The present studies examine the effects of airstream mismatch (and hence additional spanwise vorticity generation), effects of confinement of the crossflow to reduce the local equivalence ratio, and the effects of altering the geometry and position of the flameholders. NO(x) and CO emissions as well as planar laser-induced fluorescence imaging (PLIF) of seeded acetone are used to characterize injector performance and reactive flow evolution.

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

DOEpatents

Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

1982-01-01

A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Guba, G. G.

2017-11-01

This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

Full-Authority Fault-Tolerant Electronic Engine Control Systems for Variable Cycle Engines.

DTIC Science & Technology

1981-12-01

Geometry or Fuel Flow Scheduled as a Function of Engine State, i.e. FIGV = f( N1 C2 ) Closed Loop - Geometry or Fuel Flow Modulated To Maintain an Engine...Low Pressure Turbine Inlet Area (A41) Closed Loop (Integral) N2, T22 Core Stream Exhaust Nozzle Area (AJE) Closed Loop (Integral) N1 , T2 Duct Stream...to remain at the breakpoint value while low rotor speed reference ( N1 reference) is scheduled to decrease as a function of power lever angle (PLA), to

Fuel Pin Behavior Under the Slow Power Ramp Transients in the CABRI-2 Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Charpenel, Jean; Lemoine, Francette; Sato, Ikken

Slow ramp-type transient-overpower tests were performed within the framework of the international CABRI-2 experimental program. The implemented power transients of {approx}1% nominal power/s correspond to a control rod withdrawal-type accident in a liquid-metal-cooled fast breeder reactor (FBR). The analysis of the tests includes the information elements derived from the hodoscope signals, which were assessed quantitatively and supported by destructive and nondestructive posttest examinations. These tests, performed with fuels of various geometries, demonstrated the high margin to failure of such FBR fuel pins within the expected power level before the emergency reactor shutdown. At the same time, these tests performed withmore » high- and low-smear-density industrial pins led to clarification of the influence of pellet design on fuel pin behavior under high overpower condition. With the high-smear-density solid fuel pellet pin of high burnup level, the retained gaseous fission products played an important role in the solid fuel swelling, leading to clad deformation and failure at a maximum heating rate of 81 kW.m{sup -1}, which is much greater than the end-of-life (EOL) linear rating of the pin. With the low smear-density annular pellet pin, an important fuel swelling takes place, leading to degradation of the fuel thermal conductivity. This effect was detected at the power level around 73 kW.m{sup -1}, which is also much higher than the EOL value of the pin. Furthermore, the absence of clad deformation, and consequently of failure even at the power level going up to 134.7 kW.m{sup -1}, confirmed the very high margin to failure. In consequence, it was clarified that gaseous fission products have significant effects on failure threshold as well as on thermal performance during overpower condition, and such effects are significantly dependent on fuel design and power operation conditions.« less

THE MANUFACTURE OF FUEL ELEMENTS OF THE ARGONAUT TYPE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kittl, J.; Machado, R.E.; Mazza, J.A.

1958-06-10

The conditions required for the manufacture of the RA-1 Argonant type fuel elements are investigated. The fuel elements are in the form of a plate which is manufactured by the extrusion of a presintered mass of U/sub 3/O/sub 8/ (20% enriched) in an aluminum matrix. Steps in the investigation were obtention and specification of U/sub 3/O/sub 8/ and Al in powder form for testing, filling, and extrusion tests, finishing of the fuel elements, and computation of U/sub 3/O/sub 8/ content. (W.D.M.)

Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

NASA Technical Reports Server (NTRS)

Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

2013-01-01

In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

Neutronic study on conversion of SAFARI-1 to LEU silicide fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ball, G.; Pond, R.; Hanan, N.

1995-02-01

This paper marks the initial study into the technical and economic feasibility of converting the SAFARI-1 reactor in South Africa to LEU silicide fuel. Several MTR assembly geometries and LEU uranium densities have been studied and compared with MEU and HEU fuels. Two factors of primary importance for conversion of SAFARI-1 to LEU fuel are the economy of the fuel cycle and the performance of the incore and excore irradiation positions.

40 CFR 79.56 - Fuel and fuel additive grouping system.

Code of Federal Regulations, 2010 CFR

2010-07-01

... further testing under the provisions of Tier 3 or to support regulatory decisions affecting that fuel or... elements or classes of compounds other than those permitted in the base fuel for the respective fuel family... all of the following criteria: (1) Contain no elements other than carbon, hydrogen, oxygen, nitrogen...

Fracture Capabilities in Grizzly with the extended Finite Element Method (X-FEM)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dolbow, John; Zhang, Ziyu; Spencer, Benjamin

Efforts are underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). A capability was previously developed to calculate three-dimensional interaction- integrals to extract mixed-mode stress-intensity factors. This capability requires the use of a finite element mesh that conforms to the crack geometry. The eXtended Finite Element Method (X-FEM) provides a means to represent a crack geometry without explicitly fitting the finite element mesh to it. This is effected by enhancing the element kinematics to represent jump discontinuities at arbitrary locations inside ofmore » the element, as well as the incorporation of asymptotic near-tip fields to better capture crack singularities. In this work, use of only the discontinuous enrichment functions was examined to see how accurate stress intensity factors could still be calculated. This report documents the following work to enhance Grizzly’s engineering fracture capabilities by introducing arbitrary jump discontinuities for prescribed crack geometries; X-FEM Mesh Cutting in 3D: to enhance the kinematics of elements that are intersected by arbitrary crack geometries, a mesh cutting algorithm was implemented in Grizzly. The algorithm introduces new virtual nodes and creates partial elements, and then creates a new mesh connectivity; Interaction Integral Modifications: the existing code for evaluating the interaction integral in Grizzly was based on the assumption of a mesh that was fitted to the crack geometry. Modifications were made to allow for the possibility of a crack front that passes arbitrarily through the mesh; and Benchmarking for 3D Fracture: the new capabilities were benchmarked against mixed-mode three-dimensional fracture problems with known analytical solutions.« less

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tarchalski, M.; Pytel, K.; Wroblewska, M.

2015-07-01

Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to themore » qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)« less

Eigenvalue computations with the QUAD4 consistent-mass matrix

NASA Technical Reports Server (NTRS)

Butler, Thomas A.

1990-01-01

The NASTRAN user has the option of using either a lumped-mass matrix or a consistent- (coupled-) mass matrix with the QUAD4 shell finite element. At the Sixteenth NASTRAN Users' Colloquium (1988), Melvyn Marcus and associates of the David Taylor Research Center summarized a study comparing the results of the QUAD4 element with results of other NASTRAN shell elements for a cylindrical-shell modal analysis. Results of this study, in which both the lumped-and consistent-mass matrix formulations were used, implied that the consistent-mass matrix yielded poor results. In an effort to further evaluate the consistent-mass matrix, a study was performed using both a cylindrical-shell geometry and a flat-plate geometry. Modal parameters were extracted for several modes for both geometries leading to some significant conclusions. First, there do not appear to be any fundamental errors associated with the consistent-mass matrix. However, its accuracy is quite different for the two different geometries studied. The consistent-mass matrix yields better results for the flat-plate geometry and the lumped-mass matrix seems to be the better choice for cylindrical-shell geometries.

System for conversion between the boundary representation model and a constructive solid geometry model of an object

DOEpatents

Christensen, N.C.; Emery, J.D.; Smith, M.L.

1985-04-29

A system converts from the boundary representation of an object to the constructive solid geometry representation thereof. The system converts the boundary representation of the object into elemental atomic geometrical units or I-bodies which are in the shape of stock primitives or regularized intersections of stock primitives. These elemental atomic geometrical units are then represented in symbolic form. The symbolic representations of the elemental atomic geometrical units are then assembled heuristically to form a constructive solid geometry representation of the object usable for manufacturing thereof. Artificial intelligence is used to determine the best constructive solid geometry representation from the boundary representation of the object. Heuristic criteria are adapted to the manufacturing environment for which the device is to be utilized. The surface finish, tolerance, and other information associated with each surface of the boundary representation of the object are mapped onto the constructive solid geometry representation of the object to produce an enhanced solid geometry representation, particularly useful for computer-aided manufacture of the object. 19 figs.

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sell, D. A.; Baily, C. E.; Malewitz, T. J.

2016-09-01

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium aftermore » the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.« less

Fuel Cell Power Plant Initiative. Volume 2; Preliminary Design of a Fixed-Base LFP/SOFC Power System

NASA Technical Reports Server (NTRS)

Veyo, S.E.

1997-01-01

This report documents the preliminary design for a military fixed-base power system of 3 MWe nominal capacity using Westinghouse's tubular Solid Oxide Fuel Cell [SOFC] and Haldor Topsoe's logistic fuels processor [LFP]. The LFP provides to the fuel cell a methane rich sulfur free fuel stream derived from either DF-2 diesel fuel, or JP-8 turbine fuel. Fuel cells are electrochemical devices that directly convert the chemical energy contained in fuels such as hydrogen, natural gas, or coal gas into electricity at high efficiency with no intermediate heat engine or dynamo. The SOFC is distinguished from other fuel cell types by its solid state ceramic structure and its high operating temperature, nominally 1000'C. The SOFC pioneered by Westinghouse has a tubular geometry closed at one end. A power generation stack is formed by aggregating many cells in an ordered array. The Westinghouse stack design is distinguished from other fuel cell stacks by the complete absence of high integrity seals between cell elements, cells, and between stack and manifolds. Further, the reformer for natural gas [predominantly methane] and the stack are thermally and hydraulically integrated with no requirement for process water. The technical viability of combining the tubular SOFC and a logistic fuels processor was demonstrated at 27 kWe scale in a test program sponsored by the Advanced Research Projects Agency [ARPA) and carried out at the Southern California Edison's [SCE] Highgrove generating station near San Bernardino, California in 1994/95. The LFP was a breadboard design supplied by Haldor Topsoe, Inc. under subcontract to Westinghouse. The test program was completely successful. The LFP fueled the SOFC for 766 hours on JP-8 and 1555 hours of DF-2. In addition, the fuel cell operated for 3261 hours on pipeline natural gas. Over the 5582 hours of operation, the SOFC generated 118 MVVH of electricity with no perceptible degradation in performance. The LFP processed military specification JP-8 and DF-2 removing the sulfur and reforming these liquid fuels to a methane rich gaseous fuel. Results of this program are documented in a companion report titled 'Final Report-Solid Oxide Fuel Cell/ Logistic Fuels Processor 27 kWe Power System'.

Nuclear light bulb

NASA Technical Reports Server (NTRS)

Latham, Tom

1991-01-01

The nuclear light bulb engine is a closed cycle concept. The nuclear light bulb concept provides containment by keeping the nuclear fuel fluid mechanically suspended in a cylindrical geometry. Thermal heat passes through an internally cooled, fused-silica, transparent wall and heats hydrogen propellant. The seeded hydrogen propellant absorbs radiant energy and is expanded through a nozzle. Internal moderation was used in the configuration which resulted in a reduced critical density requirement. This result was supported by criticality experiments. A reference engine was designed that had seven cells and was sized to fit in what was then predicted to be the shuttle bay mass and volume limitations. There were studies done of nozzle throat cooling schemes to remove the radiant heat. Elements of the nuclear light bulb program included closed loop critical assembly tests done at Los Alamos with UF6 confined by argon buffer gas. It was shown that the fuel region could be seeded with constituents that would block UV radiation from the uranium plasma. A combination of calculations and experiments showed that internal moderation produced a critical mass reduction. Other aspects of the research are presented.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, W.E.; Trapp, T.J.

1983-06-10

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

2015-01-01

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at the time and resulted in NTREES being out of commission for a couple of months while a new stronger coil was procured. The new coil includes several additional pieces of support structure to prevent coil movement in the future. In addition, new insulating test article support components have been fabricated to prevent unexpected arcing to the test articles. Additional activities are also now underway to address ways in which the radial temperature profiles across test articles may be controlled such that they are more prototypical of what they would encounter in an operating nuclear engine. The causes of the temperature distribution problem are twofold. First, the fuel element test article is isolated in NTREES as opposed to being in the midst of many other mostly identical fuel elements in a nuclear engine. As a result, the fuel element heat flux boundary conditions in NTREES are far from adiabatic as would normally be the case in a reactor. Second, induction heating skews the power distribution such that power is preferentially deposited near the outside of the fuel element. Nuclear heating, conversely, deposits its power much more uniformly throughout the fuel element. Current studies are now looking at various schemes to adjust the amount of thermal radiation emitted from the fuel element surface so as to essentially vary the thermal boundary conditions on the test article. It is hoped that by properly adjusting the thermal boundary conditions on the fuel element test article, it may be possible to substantially correct for the inappropriate radial power distributions resulting from the induction heating so as to yield a more nearly correct temperature distribution throughout the fuel element.

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

METHOD AND APPARATUS FOR HANDLING RADIOACTIVE PRODUCTS

DOEpatents

Nicoll, D.

1959-02-24

A device is described for handling fuel elements being discharged from a nuclear reactor. The device is adapted to be disposed beneath a reactor within the storage canal for spent fuel elements. The device is comprised essentially of a cylinder pivotally mounted to a base for rotational motion between a vertical position. where the mouth of the cylinder is in the top portion of the container for receiving a fuel element discharged from a reactor into the cylinder, and a horizontal position where the mouth of the cylinder is remote from the top portion of the container and the fuel element is discharged from the cylinder into the storage canal. The device is operated by hydraulic pressure means and is provided with a means to prevent contaminated primary liquid coolant in the reactor system from entering the storage canal with the spent fuel element.

Analytical determination of thermal conductivity of W-UO2 and W-UN CERMET nuclear fuels

NASA Astrophysics Data System (ADS)

Webb, Jonathan A.; Charit, Indrajit

2012-08-01

The thermal conductivity of tungsten based CERMET fuels containing UO2 and UN fuel particles are determined as a function of particle geometry, stabilizer fraction and fuel-volume fraction, by using a combination of an analytical approach and experimental data collected from literature. Thermal conductivity is estimated using the Bruggeman-Fricke model. This study demonstrates that thermal conductivities of various CERMET fuels can be analytically predicted to values that are very close to the experimentally determined ones.

Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Shishkin, A. V.

2017-11-01

This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

Fuel cell elements with improved water handling capacity

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor); Lee, Albany (Inventor)

2001-01-01

New fuel cell components for use in liquid feed fuel cell systems are provided. The components include biplates and endplates, having a hydrophilic surface and allow high efficiency operation. Conductive elements and a wicking device also form a part of the fuel cell components of the invention.

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in thismore » paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.« less

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

DOE PAGES

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

2017-03-01

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in thismore » paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.« less

Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

NASA Astrophysics Data System (ADS)

Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

2017-04-01

Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

Boundary element based multiresolution shape optimisation in electrostatics

NASA Astrophysics Data System (ADS)

Bandara, Kosala; Cirak, Fehmi; Of, Günther; Steinbach, Olaf; Zapletal, Jan

2015-09-01

We consider the shape optimisation of high-voltage devices subject to electrostatic field equations by combining fast boundary elements with multiresolution subdivision surfaces. The geometry of the domain is described with subdivision surfaces and different resolutions of the same geometry are used for optimisation and analysis. The primal and adjoint problems are discretised with the boundary element method using a sufficiently fine control mesh. For shape optimisation the geometry is updated starting from the coarsest control mesh with increasingly finer control meshes. The multiresolution approach effectively prevents the appearance of non-physical geometry oscillations in the optimised shapes. Moreover, there is no need for mesh regeneration or smoothing during the optimisation due to the absence of a volume mesh. We present several numerical experiments and one industrial application to demonstrate the robustness and versatility of the developed approach.

Planar Solid-Oxide Fuel Cell Research and Development

DTIC Science & Technology

2013-03-28

electrolyte membrane ( PEM ) fuel cells ", Applied Surface Sei., 227 (2004) 56-72. [10] Grujicic, M., and Chittajallu, K. M., "Optimization of the...cathode geometry in polymer electrolyte membrane ( PEM ) fuel cells ", Chem. Eng. Sei., 59 (2004) 5883-5895. 36 [11] Anderson, W.K., Newman, J.C., Whitfield...M., Djilali, N, Suleman, A., "Optimization of a planar self-breathing PEM fuel cell cathode", AIAA 2006-6917, 11th AIAA/ISSMO Multidisciplinary

78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

Sensitivity and Nonlinearity of Thermoacoustic Oscillations

NASA Astrophysics Data System (ADS)

Juniper, Matthew P.; Sujith, R. I.

2018-01-01

Nine decades of rocket engine and gas turbine development have shown that thermoacoustic oscillations are difficult to predict but can usually be eliminated with relatively small ad hoc design changes. These changes can, however, be ruinously expensive to devise. This review explains why linear and nonlinear thermoacoustic behavior is so sensitive to parameters such as operating point, fuel composition, and injector geometry. It shows how nonperiodic behavior arises in experiments and simulations and discusses how fluctuations in thermoacoustic systems with turbulent reacting flow, which are usually filtered or averaged out as noise, can reveal useful information. Finally, it proposes tools to exploit this sensitivity in the future: adjoint-based sensitivity analysis to optimize passive control designs and complex systems theory to warn of impending thermoacoustic oscillations and to identify the most sensitive elements of a thermoacoustic system.

ENGINEL: A single rotor turbojet engine cycle match performance program

NASA Technical Reports Server (NTRS)

Lovell, W. A.

1977-01-01

ENGINEL is a computer program which was developed to generate the design and off-design performance of a single rotor turbojet engine with or without afterburning using a cycle match procedure. It is capable of producing engine performance over a wide range of altitudes and Mach numbers. The flexibility, of operating with a variable geometry turbine, for improved off-design fuel consumption or with a fixed geometry turbine as in conventional turbojets, has been incorporated. In addition, the option of generation engine performance with JP4, liquid hydrogen or methane as fuel is provided.

The Oldest Math Textbook.

ERIC Educational Resources Information Center

Blum, Christopher O.

2002-01-01

Asserts that when college mathematics is taught as one of the liberal arts, Euclid's Elements of Geometry, which is one of the oldest math textbooks, may still be the best, because it helps students understand what is "lovely about mathematics". Suggests that "Elements of Geometry" promotes the mental exercise that comes from…

Sandia Corporation (Albuquerque, NM)

DOEpatents

Ewsuk, Kevin G [Albuquerque, NM; Arguello, Jr., Jose G.

2006-01-31

A method of designing a primary geometry, such as for a forming die, to be used in a powder pressing application by using a combination of axisymmetric geometric shapes, transition radii, and transition spaces to simulate the geometry where the shapes can be selected from a predetermined list or menu of axisymmetric shapes and then developing a finite element mesh to represent the geometry. This mesh, along with material properties of the component to be designed and powder, is input to a standard deformation finite element code to evaluate the deformation characteristics of the component being designed. The user can develop the geometry interactively with a computer interface in minutes and execute a complete analysis of the deformation characteristics of the simulated component geometry.

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

Tag gas capsule with magnetic piercing device

DOEpatents

Nelson, Ira V.

1976-06-22

An apparatus for introducing a tag (i.e., identifying) gas into a tubular nuclear fuel element. A sealed capsule containing the tag gas is placed in the plenum in the fuel tube between the fuel and the end cap. A ferromagnetic punch having a penetrating point is slidably mounted in the plenum. By external electro-magnets, the punch may be caused to penetrate a thin rupturable end wall of the capsule and release the tag gas into the fuel element. Preferably the punch is slidably mounted within the capsule, which is in turn loaded as a sealed unit into the fuel element.

Mixing Enhancement in a Lobed Injector

NASA Technical Reports Server (NTRS)

Smith, L. L.; Majamaki, A. J.; Lam, I. T.; Delabroy, O.; Karagozian, A. R.; Marble, F. E.; Smith, O. I.

1997-01-01

An experimental investigation of the non-reactive mixing processes associated with a lobed fuel injector in a coflowing air stream is presented. The lobed fuel injector is a device which generates streamwise vorticity, producing high strain rates which can enhance the mixing of reactants while delaying ignition in a controlled manner. The lobed injectors examined in the present study consist of two corrugated plates between which a fuel surrogate, CO2, is injected into coflowing air. Acetone is seeded in the CO2 supply as a fuel marker. Comparison of two alternative lobed injector geometries is made with a straight fuel injector to determine net differences in mixing and strain fields due to streamwise vorticity generation. Planar laser-induced fluorescence (PLIF) of the seeded acetone yields two-dimensional images of the scalar concentration field at various downstream locations, from which local mixing and scalar dissipation rates are computed. It is found that the lobed injector geometry can enhance molecular mixing and create a highly strained flowfield, and that the strain rates generated by scalar energy dissipation can potentially delay ignition in a reacting flowfield.

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Nuclear reactor fuel element having improved heat transfer

DOEpatents

Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

1982-03-03

A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

Gifted Mathematicians Constructing Their Own Geometries--Changes in Knowledge and Attitude.

ERIC Educational Resources Information Center

Shillor, Irith

1997-01-01

Using Taxi-Cab Geometry (a non-Euclidean geometry program) as the starting point, 14 mathematically gifted British secondary students (ages 12-14) were asked to consider the differences between Euclidean and Non-Euclidean geometries, then to construct their own geometry and to consider the non-Euclidean elements within it. The positive effects of…

Three dimensional electrochemical simulation of solid oxide fuel cell cathode based on microstructure reconstructed by marching cubes method

NASA Astrophysics Data System (ADS)

He, An; Gong, Jiaming; Shikazono, Naoki

2018-05-01

In the present study, a model is introduced to correlate the electrochemical performance of solid oxide fuel cell (SOFC) with the 3D microstructure reconstructed by focused ion beam scanning electron microscopy (FIB-SEM) in which the solid surface is modeled by the marching cubes (MC) method. Lattice Boltzmann method (LBM) is used to solve the governing equations. In order to maintain the geometries reconstructed by the MC method, local effective diffusivities and conductivities computed based on the MC geometries are applied in each grid, and partial bounce-back scheme is applied according to the boundary predicted by the MC method. From the tortuosity factor and overpotential calculation results, it is concluded that the MC geometry drastically improves the computational accuracy by giving more precise topology information.

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

MRT fuel element inspection at Dounreay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gibson, J.

1997-08-01

To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

Time-resolved fuel injector flow characterisation based on 3D laser Doppler vibrometry

NASA Astrophysics Data System (ADS)

Crua, Cyril; Heikal, Morgan R.

2014-12-01

Hydrodynamic turbulence and cavitation are known to play a significant role in high-pressure atomizers, but the small geometries and extreme operating conditions hinder the understanding of the flow’s characteristics. Diesel internal flow experiments are generally conducted using x-ray techniques or on transparent, and often enlarged, nozzles with different orifice geometries and surface roughness to those found in production injectors. In order to enable investigations of the fuel flow inside unmodified injectors, we have developed a new experimental approach to measure time-resolved vibration spectra of diesel nozzles using a 3D laser vibrometer. The technique we propose is based on the triangulation of the vibrometer and fuel pressure transducer signals, and enables the quantitative characterisation of quasi-cyclic internal flows without requiring modifications to the injector, the working fluid, or limiting the fuel injection pressure. The vibrometer, which uses the Doppler effect to measure the velocity of a vibrating object, was used to scan injector nozzle tips during the injection event. The data were processed using a discrete Fourier transform to provide time-resolved spectra for valve-closed-orifice, minisac and microsac nozzle geometries, and injection pressures ranging from 60 to 160 MPa, hence offering unprecedented insight into cyclic cavitation and internal mechanical dynamic processes. A peak was consistently found in the spectrograms between 6 and 7.5 kHz for all nozzles and injection pressures. Further evidence of a similar spectral peak was obtained from the fuel pressure transducer and a needle lift sensor mounted into the injector body. Evidence of propagation of the nozzle oscillations to the liquid sprays was obtained by recording high-speed videos of the near-nozzle diesel jet, and computing the fast Fourier transform for a number of pixel locations at the interface of the jets. This 6-7.5 kHz frequency peak is proposed to be the natural frequency for the injector’s main internal fuel line. Other spectral peaks were found between 35 and 45 kHz for certain nozzle geometries, suggesting that these particular frequencies may be linked to nozzle dependent cavitation phenomena.

Direct carbon fuel cell and stack designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorte, Raymond J.; Oh, Tae-Sik

Disclosed are novel configurations of Direct Carbon Fuel Cells (DCFCs), which optionally comprise a liquid anode. The liquid anode comprises a molten salt/metal, preferably Sb, and a fuel, which has significant elemental carbon content (coal, bio-mass, etc.). The supply of fuel is continuously replenished in the anode. In addition, a stack configuration is suggested where combining a large number of planar or tubular fuel elements.

Nuclear breeder reactor fuel element with axial tandem stacking and getter

DOEpatents

Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

1981-01-01

A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

3D Printer-Manufacturing of Complex Geometry Elements

NASA Astrophysics Data System (ADS)

Ciubară, A.; Burlea, Ș L.; Axinte, M.; Cimpoeșu, R.; Chicet, D. L.; Manole, V.; Burlea, G.; Cimpoeșu, N.

2018-06-01

In the last 5-10 years the process of 3D printing has an incredible advanced in all the fields with a tremendous number of applications. Plastic materials exhibit highly beneficial mechanical properties while delivering complex designs impossible to achieve using conventional manufacturing. In this article the printing process (filling degree, time, complications and details finesse) of few plastic elements with complicated geometry and fine details was analyzed and comment. 3D printing offers many of the thermoplastics and industrial materials found in conventional manufacturing. The advantages and disadvantages of 3D printing for plastic parts are discussed. Time of production for an element with complex geometry, from the design to final cut, was evaluated.

Broad Specification Fuels Combustion Technology Program, Phase 2

NASA Technical Reports Server (NTRS)

Lohmann, R. P.; Jeroszko, R. A.; Kennedy, J. B.

1990-01-01

An experimental evaluation of two advanced technology combustor concepts was conducted to evolve and assess their capability for operation on broadened properties fuels. The concepts were based on the results of Phase 1 of the Broad Specification Fuel Combustor Technology Program which indicated that combustors with variable geometry or staged combustion zones had a flexibility of operation that could facilitate operation on these fuels. Emphasis in defining these concepts included the use of single pipe as opposed to duplex or staged fuels systems to avoid the risk of coking associated with the reduction in thermal stability expected in broadened properties fuels. The first concept was a variable geometry combustor in which the airflow into the primary zone could be altered through valves on the front while the second was an outgrowth of the staged Vorbix combustor, evolved under the NASA/P&W ECCP and EEE programs incorporating simplified fuel and air introduction. The results of the investigation, which involved the use of Experimental Referee Broad Specification (ERBS) fuel, indicated that in the form initially conceived, both of these combustor concepts were deficient in performance relative to many of the program goals for performance emissions. However, variations of both combustors were evaluated that incorporated features to simulate conceptual enhancement to demonstrate the long range potential of the combustor. In both cases, significant improvements relative to the program goals were observed.

Two-Dimensional Diffusion Theory Analysis of Reactivity Effects of a Fuel-Plate-Removal Experiment

NASA Technical Reports Server (NTRS)

Gotsky, Edward R.; Cusick, James P.; Bogart, Donald

1959-01-01

Two-dimensional two-group diffusion calculations were performed on the NASA reactor simulator in order to evaluate the reactivity effects of fuel plates removed successively from the center experimental fuel element of a seven- by three-element core loading at the Oak Ridge Bulk Shielding Facility. The reactivity calculations were performed by two methods: In the first, the slowing-down properties of the experimental fuel element were represented by its infinite media parameters; and, in the second, the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. The latter calculation method agreed very well with the experimented reactivity effects; the former method underestimated the experimental reactivity effects.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joshi, Jay Prakash

The objectives of this project are to calibrate the Advanced Experimental Fuel Counter (AEFC), benchmark MCNP simulations using experimental results, investigate the effects of change in fuel assembly geometry, and finally to show the boost in doubles count rates with 252Cf active soruces due to the time correlated induced fission (TCIF) effect.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Priest, N D; Richardson, R B; Edwards, G W R

2013-02-01

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

Geometry control of long-span continuous girder concrete bridge during construction through finite element model updating

NASA Astrophysics Data System (ADS)

Wu, Jie; Yan, Quan-sheng; Li, Jian; Hu, Min-yi

2016-04-01

In bridge construction, geometry control is critical to ensure that the final constructed bridge has the consistent shape as design. A common method is by predicting the deflections of the bridge during each construction phase through the associated finite element models. Therefore, the cambers of the bridge during different construction phases can be determined beforehand. These finite element models are mostly based on the design drawings and nominal material properties. However, the accuracy of these bridge models can be large due to significant uncertainties of the actual properties of the materials used in construction. Therefore, the predicted cambers may not be accurate to ensure agreement of bridge geometry with design, especially for long-span bridges. In this paper, an improved geometry control method is described, which incorporates finite element (FE) model updating during the construction process based on measured bridge deflections. A method based on the Kriging model and Latin hypercube sampling is proposed to perform the FE model updating due to its simplicity and efficiency. The proposed method has been applied to a long-span continuous girder concrete bridge during its construction. Results show that the method is effective in reducing construction error and ensuring the accuracy of the geometry of the final constructed bridge.

Nuclear fuel pin scanner

DOEpatents

Bramblett, Richard L.; Preskitt, Charles A.

1987-03-03

Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

Analysis of Fuel Vaporization, Fuel-Air Mixing, and Combustion in Integrated Mixer-Flame Holders

NASA Technical Reports Server (NTRS)

Deur, J. M.; Cline, M. C.

2004-01-01

Requirements to limit pollutant emissions from the gas turbine engines for the future High-Speed Civil Transport (HSCT) have led to consideration of various low-emission combustor concepts. One such concept is the Integrated Mixer-Flame Holder (IMFH). This report describes a series of IMFH analyses performed with KIVA-II, a multi-dimensional CFD code for problems involving sprays, turbulence, and combustion. To meet the needs of this study, KIVA-II's boundary condition and chemistry treatments are modified. The study itself examines the relationships between fuel vaporization, fuel-air mixing, and combustion. Parameters being considered include: mixer tube diameter, mixer tube length, mixer tube geometry (converging-diverging versus straight walls), air inlet velocity, air inlet swirl angle, secondary air injection (dilution holes), fuel injection velocity, fuel injection angle, number of fuel injection ports, fuel spray cone angle, and fuel droplet size. Cases are run with and without combustion to examine the variations in fuel-air mixing and potential for flashback due to the above parameters. The degree of fuel-air mixing is judged by comparing average, minimum, and maximum fuel/air ratios at the exit of the mixer tube, while flame stability is monitored by following the location of the flame front as the solution progresses from ignition to steady state. Results indicate that fuel-air mixing can be enhanced by a variety of means, the best being a combination of air inlet swirl and a converging-diverging mixer tube geometry. With the IMFH configuration utilized in the present study, flashback becomes more common as the mixer tube diameter is increased and is instigated by disturbances associated with the dilution hole flow.

Fuel loads and fuel type mapping

USGS Publications Warehouse

Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

2003-01-01

Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Model of lidar range-Doppler signatures of solid rocket fuel plumes

NASA Astrophysics Data System (ADS)

Bankman, Isaac N.; Giles, John W.; Chan, Stephen C.; Reed, Robert A.

2004-09-01

The analysis of particles produced by solid rocket motor fuels relates to two types of studies: the effect of these particles on the Earth's ozone layer, and the dynamic flight behavior of solid fuel boosters used by the NASA Space Shuttle. Since laser backscatter depends on the particle size and concentration, a lidar system can be used to analyze the particle distributions inside a solid rocket plume in flight. We present an analytical model that simulates the lidar returns from solid rocket plumes including effects of beam profile, spot size, polarization and sensing geometry. The backscatter and extinction coefficients of alumina particles are computed with the T-matrix method that can address non-spherical particles. The outputs of the model include time-resolved return pulses and range-Doppler signatures. Presented examples illustrate the effects of sensing geometry.

Dynamic interactions between hypersonic vehicle aerodynamics and propulsion system performance

NASA Technical Reports Server (NTRS)

Flandro, G. A.; Roach, R. L.; Buschek, H.

1992-01-01

Described here is the development of a flexible simulation model for scramjet hypersonic propulsion systems. The primary goal is determination of sensitivity of the thrust vector and other system parameters to angle of attack changes of the vehicle. Such information is crucial in design and analysis of control system performance for hypersonic vehicles. The code is also intended to be a key element in carrying out dynamic interaction studies involving the influence of vehicle vibrations on propulsion system/control system coupling and flight stability. Simple models are employed to represent the various processes comprising the propulsion system. A method of characteristics (MOC) approach is used to solve the forebody and external nozzle flow fields. This results in a very fast computational algorithm capable of carrying out the vast number of simulation computations needed in guidance, stability, and control studies. The three-dimensional fore- and aft body (nozzle) geometry is characterized by the centerline profiles as represented by a series of coordinate points and body cross-section curvature. The engine module geometry is represented by an adjustable vertical grid to accommodate variations of the field parameters throughout the inlet and combustor. The scramjet inlet is modeled as a two-dimensional supersonic flow containing adjustable sidewall wedges and multiple fuel injection struts. The inlet geometry including the sidewall wedge angles, the number of injection struts, their sweepback relative to the vehicle reference line, and strut cross-section are user selectable. Combustion is currently represented by a Rayleigh line calculation including corrections for variable gas properties; improved models are being developed for this important element of the propulsion flow field. The program generates (1) variation of thrust magnitude and direction with angle of attack, (2) pitching moment and line of action of the thrust vector, (3) pressure and temperature distributions throughout the system, and (4) performance parameters such as thrust coefficient, specific impulse, mass flow rates, and equivalence ratio. Preliminary results are in good agreement with available performance data for systems resembling the NASP vehicle configuration.

Urban geometry effects on the microclimate of streets near the river

NASA Astrophysics Data System (ADS)

Park, C.; Lee, D. K.; Asawa, T.; Murakami, A.; Kim, H. G.; Jeong, W.

2016-12-01

These days, many urbans have undergone urban heat islands (UHIs) which mean urban areas have higher temperature than surrounding areas. Because of UHIs, extreme heat waves and tropical nights threaten citizen's health. For reducing UHI effects, UHIs mitigation strategies related with urban planning (cooling elements, urban geometry, and urban materials) are studied in many fields. Especially, cooling elements such as vegetation and water are important for making sustainable cities. However, while the each effects of cooling elements are known, combination effects between urban geometry and elements are not clearly defined. To determine which urban geometry is effective to decrease air temperature, we analyzed urban geometry effects on the microclimate of streets near the river. The urban streets located in commercial area near the Cheonggye River and they had different aspect ratio. For doing this, we used field measurement. Field measurement were takin in the sunny days of July and August 2016. Air temperature measurement used T-type thermometers with cylindrical tube and ventilation fan. Wind measurements used kestrel 4500 for wind speed and wind direction. We measured air temperature in the screen level and repeated three times an hour. Measurement results showed that large aspect ratio street show lower air temperature than small ones. And they received cooling effect farther than others from the river. This is because that cooling air flows well in the wide road streets. Large aspect ratio streets are effective to receive river's cooling air. These results can be used in the urban geometry planning near the river.

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Determination of trace elements in automotive fuels by filter furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Anselmi, Anna; Tittarelli, Paolo; Katskov, Dmitri A.

2002-03-01

The determination of Cd, Cr, Cu, Pb and Ni was performed in gasoline and diesel fuel samples by electrothermal atomic absorption spectrometry using the Transverse Heated Filter Atomizer (THFA). Thermal conditions were experimentally defined for the investigated elements. The elements were analyzed without addition of chemical modifiers, using organometallic standards for the calibration. Forty-microliter samples were injected into the THFA. Gasoline samples were analyzed directly, while diesel fuel samples were diluted 1:4 with n-heptane. The following characteristic masses were obtained: 0.8 pg Cd, 6.4 pg Cr, 12 pg Cu, 17 pg Pb and 27 pg Ni. The limits of determination for gasoline samples were 0.13 μg/kg Cd, 0.4 μg/kg Cr, 0.9 μg/kg Cu, 1.5 μg/kg Pb and 2.5 μg/kg Ni. The corresponding limit of determination for diesel fuel samples was approximately four times higher for all elements. The element recovery was performed using the addition of organometallic compounds to gasoline and diesel fuel samples and was between 85 and 105% for all elements investigated.

FUEL ELEMENT

DOEpatents

Bean, R.W.

1963-11-19

A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Hofman, G.L.

1997-12-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data.

Reduced size fuel cell for portable applications

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor); Clara, Filiberto (Inventor); Frank, Harvey A. (Inventor)

2004-01-01

A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

Coolant mass flow equalizer for nuclear fuel

DOEpatents

Betten, Paul R.

1978-01-01

The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

Interior flow and near-nozzle spray development in a marine-engine diesel fuel injector

NASA Astrophysics Data System (ADS)

Hult, J.; Simmank, P.; Matlok, S.; Mayer, S.; Falgout, Z.; Linne, M.

2016-04-01

A consolidated effort at optically characterising flow patterns, in-nozzle cavitation, and near-nozzle jet structure of a marine diesel fuel injector is presented. A combination of several optical techniques was employed to fully transparent injector models, compound metal-glass and full metal injectors. They were all based on a common real-scale dual nozzle hole geometry for a marine two-stroke diesel engine. In a stationary flow rig, flow velocities in the sac-volume and nozzle holes were measured using PIV, and in-nozzle cavitation visualized using high-resolution shadowgraphs. The effect of varying cavitation number was studied and results compared to CFD predictions. In-nozzle cavitation and near-nozzle jet structure during transient operation were visualized simultaneously, using high-speed imaging in an atmospheric pressure spray rig. Near-nozzle spray formation was investigated using ballistic imaging. Finally, the injector geometry was tested on a full-scale marine diesel engine, where the dynamics of near-nozzle jet development was visualized using high-speed shadowgraphy. The range of studies focused on a single common geometry allows a comprehensive survey of phenomena ranging from first inception of cavitation under well-controlled flow conditions to fuel jet structure at real engine conditions.

Improved nuclear fuel assembly grid spacer

DOEpatents

Marshall, John; Kaplan, Samuel

1977-01-01

An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

DOEpatents

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

JACKETED FUEL ELEMENT

DOEpatents

Wigner, E.P.; Szilard, L.; Creutz, E.C.

1959-02-01

These fuel elements are comprised of a homogeneous metallic uranium body completely enclosed and sealed in an aluminum cover. The uranium body and aluminum cover are bonded together by a layer of zinc located between them. The bonding layer serves to improve transfer of heat, provides an additional protection against corrosion of the uranium by the coolant, and also localizes any possible corrosion by preventing travel of corrosive material along the surface of the fuel element.

Comprehensive Fuel Spray Modeling and Impacts on Chamber Acoustics in Combustion Dynamics Simulations

DTIC Science & Technology

2013-05-01

multiple swirler configurations and fuel injector locations at atmospheric pressure con- ditions. Both single-element and multiple-element LDI...the swirl number, Reynolds’ number and injector location in the LDI element. Besides the multi-phase flow characteristics, several experimen- tal...region downstream of the fuel injector on account of a sta- ble and compact precessing vortex core. Recent ex- periments conducted by the Purdue group have

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

DOEpatents

Flint, O.

1961-01-10

Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruggles, A.E.

1990-10-12

The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results aremore » related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.« less

The sensitivity of biological finite element models to the resolution of surface geometry: a case study of crocodilian crania

PubMed Central

Evans, Alistair R.; McHenry, Colin R.

2015-01-01

The reliability of finite element analysis (FEA) in biomechanical investigations depends upon understanding the influence of model assumptions. In producing finite element models, surface mesh resolution is influenced by the resolution of input geometry, and influences the resolution of the ensuing solid mesh used for numerical analysis. Despite a large number of studies incorporating sensitivity studies of the effects of solid mesh resolution there has not yet been any investigation into the effect of surface mesh resolution upon results in a comparative context. Here we use a dataset of crocodile crania to examine the effects of surface resolution on FEA results in a comparative context. Seven high-resolution surface meshes were each down-sampled to varying degrees while keeping the resulting number of solid elements constant. These models were then subjected to bite and shake load cases using finite element analysis. The results show that incremental decreases in surface resolution can result in fluctuations in strain magnitudes, but that it is possible to obtain stable results using lower resolution surface in a comparative FEA study. As surface mesh resolution links input geometry with the resulting solid mesh, the implication of these results is that low resolution input geometry and solid meshes may provide valid results in a comparative context. PMID:26056620

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Potter, David Charles; Taylor, Craig Michael; Coons, James Elmer

The percent void of the Fort Saint Vrain (FSV) material is estimated to be 21.1% based on the volume of the gap at the top of the drums, the volume of the coolant channels in the FSV fuel element, and the volume of the fuel handling channel in the FSV fuel element.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Gurinsky, D.H.; Powell, R.W.; Fox, M.

1959-11-24

A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

Space reactor fuel element testing in upgraded TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

DOEpatents

Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

1963-09-01

A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

Finite element area and line integral transforms for generalization of aperture function and geometry in Kirchhoff scalar diffraction theory

NASA Astrophysics Data System (ADS)

Kraus, Hal G.

1993-02-01

Two finite element-based methods for calculating Fresnel region and near-field region intensities resulting from diffraction of light by two-dimensional apertures are presented. The first is derived using the Kirchhoff area diffraction integral and the second is derived using a displaced vector potential to achieve a line integral transformation. The specific form of each of these formulations is presented for incident spherical waves and for Gaussian laser beams. The geometry of the two-dimensional diffracting aperture(s) is based on biquadratic isoparametric elements, which are used to define apertures of complex geometry. These elements are also used to build complex amplitude and phase functions across the aperture(s), which may be of continuous or discontinuous form. The finite element transform integrals are accurately and efficiently integrated numerically using Gaussian quadrature. The power of these methods is illustrated in several examples which include secondary obstructions, secondary spider supports, multiple mirror arrays, synthetic aperture arrays, apertures covered by screens, apodization, phase plates, and off-axis apertures. Typically, the finite element line integral transform results in significant gains in computational efficiency over the finite element Kirchhoff transform method, but is also subject to some loss in generality.

Free-jet Testing of a REST Scramjet at Off-Design Conditions

NASA Technical Reports Server (NTRS)

Smart, Michael K.; Ruf, Edward G.

2006-01-01

Scramjet flowpaths employing elliptical combustors have the potential to improve structural efficiency and performance relative to those using planar geometries. NASA Langley has developed a scramjet flowpath integrated into a lifting body vehicle, while transitioning from a rectangular capture area to both an elliptical throat and combustor. This Rectangular-to-Elliptical Shape Transition (REST) scramjet, has a design point of Mach 7.1, and is intended to operate with fixed-geometry between Mach 4.5 and 8.0. This paper describes initial free-jet testing of the heat-sink REST scramjet engine model at conditions simulating Mach 5.3 flight. Combustion of gaseous hydrogen fuel at equivalence ratios between 0.5 and 1.5 generated robust performance after ignition with a silane-hydrogen pilot. Facility model interactions were experienced for fuel equivalence ratios above 1.1, yet despite this, the flowpath was not unstarted by fuel addition at the Mach 5.3 test condition. Combustion tests at reduced stagnation enthalpy indicated that the engine self-started following termination of the fuel injection. Engine data is presented for the largest fuel equivalence ratio tested without facility interaction. These results indicate that this class of three-dimensional scramjet engine operates successfully at off-design conditions.

Fuel shipment experience, fuel movements from the BMI-1 transport cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bauer, Thomas L.; Krause, Michael G

1986-07-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including severalmore » factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.« less

The Finite Element Modelling and Dynamic Characteristics Analysis about One Kind of Armoured Vehicles’ Fuel Tanks

NASA Astrophysics Data System (ADS)

Gao, Yang; Ge, Zhishang; Zhai, Weihao; Tan, Shiwang; Zhang, Feng

2018-01-01

The static and dynamic characteristics of fuel tank are studied for the armoured vehicle in this paper. The CATIA software is applied to build the CAD model of the armoured vehicles’ fuel tank, and the finite element model is established in ANSYS Workbench. The finite element method is carried out to analyze the static and dynamic mechanical properties of the fuel tank, and the first six orders of mode shapes and their frequencies are also computed and given in the paper, then the stress distribution diagram and the high stress areas are obtained. The results of the research provide some references to the fuel tanks’ design improvement, and give some guidance for the installation of the fuel tanks on armoured vehicles, and help to improve the properties and the service life of this kind of armoured vehicles’ fuel tanks.

Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

NASA Astrophysics Data System (ADS)

Lewis, operating defective fuel B. J.; Thompson, W. T.; Akbari, F.; Thompson, D. M.; Thurgood, C.; Higgs, J.

2004-07-01

A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor.

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

DOEpatents

Handwerk, J.H.; Noland, R.A.; Walker, D.E.

1957-09-10

In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE PAGES

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; ...

2016-10-01

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

NUCLEAR REACTOR

DOEpatents

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

JACKETED FISSIONABLE MEMBER

DOEpatents

Boller, E.R.; Robinson, J.W.

1960-09-13

A fuel element design for a nuclear reactor is presented. The fuel element comprises a cylindrical fuel body having a portion of smaller diameter at each end thereof with an annular flange at the extreme ends of these portions of smaller diameter. An end cap fits over the ends of the fuel body and has an internal annular groove adapted to receive the flange. The fuel body and end caps are disposed in a cup-shaped jacket, a closure disc completing the enclosure of the fuel body, and tht caps are bonded over their entire periphery to the jacket.

U-Mo Plate Blister Anneal Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francine J. Rice; Daniel M. Wachs; Adam B. Robinson

2010-10-01

Blister thresholds in fuel elements have been a longstanding performance parameter for fuel elements of all types. This behavior has yet to be fully defined for the RERTR U-Mo fuel types. Blister anneal studies that began in 2007 have been expanded to include plates from more recent RERTR experiments. Preliminary data presented in this report encompasses the early generations of the U-Mo fuel systems and the most recent but still developing fuel system. Included is an overview of relevant dispersion fuel systems for the purposes of comparison.

A comparative study on different methods of automatic mesh generation of human femurs.

PubMed

Viceconti, M; Bellingeri, L; Cristofolini, L; Toni, A

1998-01-01

The aim of this study was to evaluate comparatively five methods for automating mesh generation (AMG) when used to mesh a human femur. The five AMG methods considered were: mapped mesh, which provides hexahedral elements through a direct mapping of the element onto the geometry; tetra mesh, which generates tetrahedral elements from a solid model of the object geometry; voxel mesh which builds cubic 8-node elements directly from CT images; and hexa mesh that automatically generated hexahedral elements from a surface definition of the femur geometry. The various methods were tested against two reference models: a simplified geometric model and a proximal femur model. The first model was useful to assess the inherent accuracy of the meshes created by the AMG methods, since an analytical solution was available for the elastic problem of the simplified geometric model. The femur model was used to test the AMG methods in a more realistic condition. The femoral geometry was derived from a reference model (the "standardized femur") and the finite element analyses predictions were compared to experimental measurements. All methods were evaluated in terms of human and computer effort needed to carry out the complete analysis, and in terms of accuracy. The comparison demonstrated that each tested method deserves attention and may be the best for specific situations. The mapped AMG method requires a significant human effort but is very accurate and it allows a tight control of the mesh structure. The tetra AMG method requires a solid model of the object to be analysed but is widely available and accurate. The hexa AMG method requires a significant computer effort but can also be used on polygonal models and is very accurate. The voxel AMG method requires a huge number of elements to reach an accuracy comparable to that of the other methods, but it does not require any pre-processing of the CT dataset to extract the geometry and in some cases may be the only viable solution.

A small, 1400 deg Kelvin, reactor for Brayton space power systems

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Triplett, C.E.

1996-12-01

This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan`s investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the formmore » Nu = C(Ra){sup n}, where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan`s aligned array results and to other studies of natural convection in horizontal tube arrays.« less

Development of an LS-DYNA Model of an ATR42-300 Aircraft for Crash Simulation

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Fasanella, Edwin L.

2004-01-01

This paper describes the development of an LS-DYNA simulation of a vertical drop test of an ATR42-300 twin-turboprop high-wing commuter-class airplane. A 30-ft/s drop test of this aircraft was performed onto a concrete impact surface at the FAA Technical Center on July 30, 2003. The purpose of the test was to evaluate the structural response of a commuter-class aircraft when subjected to a severe, but survivable, impact. The aircraft was configured with crew and passenger seats, anthropomorphic test dummies, forward and aft luggage, instrumentation, and onboard data acquisition systems. The wings were filled with approximately 8,700 lb. of water to represent the fuel and the aircraft weighed a total of 33,200 lb. The model, which consisted of 57,643 nodes and 62,979 elements, was developed from direct measurements of the airframe geometry, over a period of approximately 8 months. The seats, dummies, luggage, fuel, and other ballast were represented using concentrated masses. Comparisons were made of the structural deformation and failure behavior of the airframe, as well as selected acceleration time history responses.

Cavitation Inside High-Pressure Optically Transparent Fuel Injector Nozzles

NASA Astrophysics Data System (ADS)

Falgout, Z.; Linne, M.

2015-12-01

Nozzle-orifice flow and cavitation have an important effect on primary breakup of sprays. For this reason, a number of studies in recent years have used injectors with optically transparent nozzles so that orifice flow cavitation can be examined directly. Many of these studies use injection pressures scaled down from realistic injection pressures used in modern fuel injectors, and so the geometry must be scaled up so that the Reynolds number can be matched with the industrial applications of interest. A relatively small number of studies have shown results at or near the injection pressures used in real systems. Unfortunately, neither the specifics of the design of the optical nozzle nor the design methodology used is explained in detail in these papers. Here, a methodology demonstrating how to prevent failure of a finished design made from commonly used optically transparent materials will be explained in detail, and a description of a new design for transparent nozzles which minimizes size and cost will be shown. The design methodology combines Finite Element Analysis with relevant materials science to evaluate the potential for failure of the finished assembly. Finally, test results imaging a cavitating flow at elevated pressures are presented.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Petrucco, Louis Jacob (Inventor); Daum, Wolfgang (Inventor)

2005-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

2003-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

1999-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Local Burn-Up Effects in the NBSR Fuel Element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown N. R.; Hanson A.; Diamond, D.

2013-01-31

This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peakingmore » relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.« less

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

DOEpatents

Draley, J.E.; Ruther, W.E.

1959-01-27

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

Flow visualization studies of transverse fuel injection patterns in a nonreacting Mach 2 combustor

NASA Technical Reports Server (NTRS)

Mcdaniel, J. C.

1987-01-01

Planar visualization images are recorded of transverse jet mixing in a supersonic combustor flowfield, without chemical reaction, using laser-induced fluorescence from iodine molecules. Digital image processing and three-dimensional display enable complete representations of fuel penetration boundary and shock surfaces corresponding to several injection geometries and pressures.

NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

DOEpatents

McCorkle, W.H.

1960-02-23

BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

Dynamics and control of state-dependent networks for probing genomic organization

PubMed Central

Rajapakse, Indika; Groudine, Mark; Mesbahi, Mehran

2011-01-01

A state-dependent dynamic network is a collection of elements that interact through a network, whose geometry evolves as the state of the elements changes over time. The genome is an intriguing example of a state-dependent network, where chromosomal geometry directly relates to genomic activity, which in turn strongly correlates with geometry. Here we examine various aspects of a genomic state-dependent dynamic network. In particular, we elaborate on one of the important ramifications of viewing genomic networks as being state-dependent, namely, their controllability during processes of genomic reorganization such as in cell differentiation. PMID:21911407

FUEL ELEMENT

DOEpatents

Fortescue, P.; Zumwalt, L.R.

1961-11-28

A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

Atomization and Dispersion of a Liquid Jet Injected Into a Crossflow of Air

NASA Technical Reports Server (NTRS)

Seay, J. E.; Samuelson, G. S.

1996-01-01

In recent years, environmental regulations have become more stringent, requiring lower emissions of mainly nitrogen oxides (NOx), as well as carbon monoxide (CO) and unburned hydrocarbons (UHC). These regulations have forced the gas turbine industry to examine non-conventional combustion strategies, such as the lean burn approach. The reasoning behind operating under lean conditions is to maintain the temperature of combustion near and below temperatures required for the formation of thermal nitric oxide (NO). To be successful, however, the lean processes require careful preparation of the fuel/air mixture to preclude formation of either locally rich reaction zones, which may give rise to NO formation, or locally lean reaction zones, which may give rise to inefficient fuel processing. As a result fuel preparation is crucial to the development and success of new aeroengine combustor technologies. A key element of the fuel preparation process is the fuel nozzle. As nozzle technologies have developed, airblast atomization has been adopted for both industrial and aircraft gas turbine applications. However, the majority of the work to date has focused on prefilming nozzles, which despite their complexity and high cost have become an industry standard for conventional combustion strategies. It is likely that the new strategies required to meet future emissions goals will utilize novel fuel injector approaches, such as radial injection. This thesis proposes and demonstrates an experiment to examine, on a mechanistic level (i.e., the physics of the action), the processes associated with the atomization, evaporation, and dispersion of a liquid jet introduced, from a radial, plain-jet airblast injector, into a crossflow of air. This understanding requires the knowledge not only of what factors influence atomization, but also the underlying mechanism associated with liquid breakup and dispersion. The experimental data acquired identify conditions and geometries for improved performance of radial airblast injectors.

FUEL ELEMENT CONSTRUCTION

DOEpatents

Simnad, M.T.

1961-08-15

A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

The Guardian: The Source for Antiterrorism Information. Volume 9, Number 1, April 2007

DTIC Science & Technology

2007-04-01

the fuel in these research reactors is generally not highly radioactive . Unlike the fuel rods in a nuclear power plant, these fuel elements would...NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING ORGANIZATION NAME(S) AND...practices and lessons learned. In addition, we will include Service and issue-specific breakout sessions that will focus on critical AT program elements

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

2011-01-01

Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

Compact Fuel Element Environment Test

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

2012-01-01

Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

Dart model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Hofman, G.L.

1997-06-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density.more » Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.« less

RERTR-12 Post-irradiation Examination Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, Francine; Williams, Walter; Robinson, Adam

2015-02-01

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-propertymore » changes of the fuel and cladding.« less

IRRADIATION METHOD AND APPARATUS

DOEpatents

Cabell, C.P.

1962-12-18

A method and apparatus are described for changing fuel bodies into a process tube of a reactor. According to this method fresh fuel elements are introduced into one end of the tube forcing used fuel elements out the other end. When sufficient fuel has been discharged, a reel and tape arrangement is employed to pull the column of bodies back into the center of the tube. Due provision is made for providing shielding in the tube. (AEC)

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

1992-01-01

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

Using Mobile Devices to Enhance the Interactive Learning for Spatial Geometry

ERIC Educational Resources Information Center

Chang, Kuo-En; Wu, Lin-Jung; Lai, Shing-Chuang; Sung, Yao-Ting

2016-01-01

The purpose of this research is to develop a hands-on spatial geometry learning system to facilitate the learning of geometry. The development of this system was based on Duval's four critical elements of geometric learning: perceptual apprehension, sequential apprehension, operative apprehension, and discursive apprehension. The system offers…

MELCOR model for an experimental 17x17 spent fuel PWR assembly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey

2010-11-01

A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

Definition of NASTRAN sets by use of parametric geometry

NASA Technical Reports Server (NTRS)

Baughn, Terry V.; Tiv, Mehran

1989-01-01

Many finite element preprocessors describe finite element model geometry with points, lines, surfaces and volumes. One method for describing these basic geometric entities is by use of parametric cubics which are useful for representing complex shapes. The lines, surfaces and volumes may be discretized for follow on finite element analysis. The ability to limit or selectively recover results from the finite element model is extremely important to the analyst. Equally important is the ability to easily apply boundary conditions. Although graphical preprocessors have made these tasks easier, model complexity may not lend itself to easily identify a group of grid points desired for data recovery or application of constraints. A methodology is presented which makes use of the assignment of grid point locations in parametric coordinates. The parametric coordinates provide a convenient ordering of the grid point locations and a method for retrieving the grid point ID's from the parent geometry. The selected grid points may then be used for the generation of the appropriate set and constraint cards.

Non-homogeneous hybrid rocket fuel for enhanced regression rates utilizing partial entrainment

NASA Astrophysics Data System (ADS)

Boronowsky, Kenny

A concept was developed and tested to enhance the performance and regression rate of hydroxyl terminated polybutadiene (HTPB), a commonly used hybrid rocket fuel. By adding small nodules of paraffin into the HTPB fuel, a non-homogeneous mixture was created resulting in increased regression rates. The goal was to develop a fuel with a simplified single core geometry and a tailorable regression rate. The new fuel would benefit from the structural stability of HTPB yet not suffer from the large void fraction representative of typical HTPB core geometries. Regression rates were compared between traditional HTPB single core grains, 85% HTPB mixed with 15% (by weight) paraffin cores, 70% HTPB mixed with 30% paraffin cores, and plain paraffin single core grains. Each fuel combination was tested at oxidizer flow rates, ranging from 0.9 - 3.3 g/s of gaseous oxygen, in a small scale hybrid test rocket and average regression rates were measured. While large uncertainties were present in the experimental setup, the overall data showed that the regression rate was enhanced as paraffin concentration increased. While further testing would be required at larger scales of interest, the trends are encouraging. Inclusion of paraffin nodules in the HTPB grain may produce a greater advantage than other more noxious additives in current use. In addition, it may lead to safer rocket motors with higher integrated thrust due to the decreased void fraction.

Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

DOEpatents

Cowell, G.K.; DiGuiseppe, C.P.

1982-10-28

A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

Thermal stress analysis of symmetric shells subjected to asymmetric thermal loads

NASA Technical Reports Server (NTRS)

Negaard, G. R.

1980-01-01

The performance of the NASTRAN level 16.0 axisymmetric solid elements when subjected to both symmetric and asymmetric thermal loading was investigated. A ceramic radome was modeled using both the CTRAPRG and the CTRAPAX elements. The thermal loading applied contained severe gradients through the thickness of the shell. Both elements were found to be more sensitive to the effect of the thermal gradient than to the aspect ratio of the elements. Analysis using the CTRAPAX element predicted much higher thermal stresses than the analysis using the CTRAPRG element, prompting studies of models for which theoretical solutions could be calculated. It was found that the CTRAPRG element solutions were satisfactory, but that the CTRAPAX element was very geometry dependent. This element produced erroneous results if the geometry was allowed to vary from a rectangular cross-section. The most satisfactory solution found for this type of problem was to model a small segment of a symmetric structure with isoparametric solid elements and apply the cyclic symmetry option in NASTRAN.

Analysis of regenerated single-shaft ceramic gas-turbine engines and resulting fuel economy in a compact car

NASA Technical Reports Server (NTRS)

Klann, J. L.; Tew, R. C., Jr.

1977-01-01

Ranges in design and off-design operating conditions of an advanced gas turbine and their effects on fuel economy were analyzed. The assumed engine incorporated a single stage radial flow turbine and compressor with fixed geometry. Fuel economies were calculated over the composite driving cycle with gasoline as the fuel. At a constant turbine-inlet temperature, with a regenerator sized for a full power effectiveness the best fuel economies ranged from 11.1 to 10.2 km/liter (26.2 to 22.5 mpg) for full power turbine tip speeds of 770 to 488m/sec (2530 to 1600ft/sec), respectively.

PREIRRADIATION MEASUREMENTS OF PIQUA FUEL ELEMENTS NO. P-1111, P-1113, P- 1114, AND P-1120

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hubbell, H.J.

1962-11-01

Results of preirradiation measurements and tests performed during the processing and assembly of the individual fuel cylinders contained in Piqua Fuel Elements No. P-1111, P-1113, P-1114, and P-1120 are presented. A description of the techniques and equipment used in obtaining the data is also included. (auth)

Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pitner, A.L.

1998-09-11

Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading.

Photographic combustion characterization of LOX/hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1979-01-01

Single element injectors and two fuels were tested with the aim of photographically characterizing observed combustion phenomena. The three injectors tested were the O-F-O triplet, the transverse like on like (TLOL), and the rectangular unlike doublet (RUD). The fuels tested were RP-1 and propane. The hot firings were conducted in a specifically constructed chamber fitted with quartz windows for photographically viewing the impingement spray field. All LOX/HC testing demonstrated coking with the RP-1 fuel leaving far more soot than the propane fuel. No fuel freezing or popping was experienced under the test conditions evaluated. Carbon particle emission and combustion light brilliance increased with Pc for both fuels although RP-1 was far more energetic in this respect. The RSS phenomena appear to be present in the high Pc tests as evidenced by striations in the spray pattern and by separate fuel rich and oxidizer rich areas. The RUD element was also tested as a fuel rich gas generator element by switching the propellant circuits. Excessive sooting occurred at this low mixture ratio (0.55), precluding photographic data.

NEUTRONIC REACTOR CONTROL ELEMENT

DOEpatents

Newson, H.W.

1960-09-13

A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

Reduce beam hardening artifacts of polychromatic X-ray computed tomography by an iterative approximation approach.

PubMed

Shi, Hongli; Yang, Zhi; Luo, Shuqian

2017-01-01

The beam hardening artifact is one of most important modalities of metal artifact for polychromatic X-ray computed tomography (CT), which can impair the image quality seriously. An iterative approach is proposed to reduce beam hardening artifact caused by metallic components in polychromatic X-ray CT. According to Lambert-Beer law, the (detected) projections can be expressed as monotonic nonlinear functions of element geometry projections, which are the theoretical projections produced only by the pixel intensities (image grayscale) of certain element (component). With help of a prior knowledge on spectrum distribution of X-ray beam source and energy-dependent attenuation coefficients, the functions have explicit expressions. Newton-Raphson algorithm is employed to solve the functions. The solutions are named as the synthetical geometry projections, which are the nearly linear weighted sum of element geometry projections with respect to mean of each attenuation coefficient. In this process, the attenuation coefficients are modified to make Newton-Raphson iterative functions satisfy the convergence conditions of fixed pointed iteration(FPI) so that the solutions will approach the true synthetical geometry projections stably. The underlying images are obtained using the projections by general reconstruction algorithms such as the filtered back projection (FBP). The image gray values are adjusted according to the attenuation coefficient means to obtain proper CT numbers. Several examples demonstrate the proposed approach is efficient in reducing beam hardening artifacts and has satisfactory performance in the term of some general criteria. In a simulation example, the normalized root mean square difference (NRMSD) can be reduced 17.52% compared to a newest algorithm. Since the element geometry projections are free from the effect of beam hardening, the nearly linear weighted sum of them, the synthetical geometry projections, are almost free from the effect of beam hardening. By working out the synthetical geometry projections, the proposed approach becomes quite efficient in reducing beam hardening artifacts.

3D Simulations of the ``Keyhole'' Hohlraum for Shock Timing on NIF

NASA Astrophysics Data System (ADS)

Robey, H. F.; Marinak, M. M.; Munro, D. H.; Jones, O. S.

2007-11-01

Ignition implosions planned for the National Ignition Facility (NIF) require a pulse shape with a carefully designed series of steps, which launch a series of shocks through the ablator and DT fuel. The relative timing of these shocks must be tuned to better than +/- 100ps to maintain the DT fuel on a sufficiently low adiabat. To meet these requirements, pre-ignition tuning experiments using a modified hohlraum geometry are being planned. This modified geometry, known as the ``keyhole'' hohlraum, adds a re-entrant gold cone, which passes through the hohlraum and capsule walls, to provide an optical line-of-sight to directly measure the shocks as they break out of the ablator. In order to assess the surrogacy of this modified geometry, 3D simulations using HYDRA [1] have been performed. The drive conditions and the resulting effect on shock timing in the keyhole hohlraum will be compared with the corresponding results for the standard ignition hohlraum. [1] M.M. Marinak, et al., Phys. Plasmas 8, 2275 (2001).

Effect of fuel and nozzle geometry on the off-axis oscillation of needle in diesel injectors using high-speed X-ray phase contrast imaging

NASA Astrophysics Data System (ADS)

Zhang, X.; Liu, J.; Wang, J.

2016-05-01

The diesel spray characteristics are strongly influenced by the flow dynamics inside the injector nozzle. Moreover, the off-axis oscillation of needle could lead to variation of orifice flow in the nozzle. In this paper, the needle oscillation was investigated using high-speed X-ray phase contrast imaging and quantitative image processing. The effects of fuel, injection pressure and nozzle geometry on the needle oscillation were analyzed. The results showed that the vertical and horizontal oscillation of needle was independent on the injection pressure. The maximum oscillation range of 14μ m was found. Biodiesel application slightly decreased the needle oscillation due to high viscosity. The needle oscillation range increased generally with increasing hole number. The larger needle oscillation in multi-hole injectors was dominated by the geometry problem or production issue at lower needle lift. In addition, the influence of needle oscillation on the spray morphology was also discussed.

ELECTROLYTIC SEPARATION PROCESS AND APPARATUS

DOEpatents

McLain, M.E. Jr.; Roberts, M.W.

1962-03-01

A method is given for dissolving stainless steel-c lad fuel elements in dilute acids such as half normal sulfuric acid. The fuel element is made the anode in a Y-shaped electrolytic cell which has a flowing mercury cathode; the stainless steel elements are entrained in the mercury and stripped therefrom by a continuous process. (AEC)

Nuclear breeder reactor fuel element with silicon carbide getter

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1987-01-01

An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Minimizing Actuator-Induced Residual Error in Active Space Telescope Primary Mirrors

DTIC Science & Technology

2010-09-01

actuator geometry, and rib-to-facesheet intersection geometry are exploited to achieve improved performance in silicon carbide ( SiC ) mirrors . A...are exploited to achieve improved performance in silicon carbide ( SiC ) mirrors . A parametric finite element model is used to explore the trade space...MOST) finite element model. The move to lightweight actively-controlled silicon carbide ( SiC ) mirrors is traced back to previous generations of space

Liquid fuel injection elements for rocket engines

NASA Technical Reports Server (NTRS)

Cox, George B., Jr. (Inventor)

1993-01-01

Thrust chambers for liquid propellant rocket engines include three principal components. One of these components is an injector which contains a plurality of injection elements to meter the flow of propellants at a predetermined rate, and fuel to oxidizer mixture ratio, to introduce the mixture into the combustion chamber, and to cause them to be atomized within the combustion chamber so that even combustion takes place. Evolving from these injectors are tube injectors. These tube injectors have injection elements for injecting the oxidizer into the combustion chamber. The oxidizer and fuel must be metered at predetermined rates and mixture ratios in order to mix them within the combustion chamber so that combustion takes place smoothly and completely. Hence tube injectors are subject to improvement. An injection element for a liquid propellant rocket engine of the bipropellant type is provided which includes tangential fuel metering orifices, and a plurality of oxidizer tube injection elements whose injection tubes are also provided with tangential oxidizer entry slots and internal reed valves.

FLUID MODERATED REACTOR

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

NASA Technical Reports Server (NTRS)

Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

2013-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

Improved gas tagging and cover gas combination for nuclear reactor

DOEpatents

Gross, K.C.; Laug, M.T.

1983-09-26

The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

Review of Rover fuel element protective coating development at Los Alamos

NASA Technical Reports Server (NTRS)

Wallace, Terry C.

1991-01-01

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

A Coupling Strategy of FEM and BEM for the Solution of a 3D Industrial Crack Problem

NASA Astrophysics Data System (ADS)

Kouitat Njiwa, Richard; Taha Niane, Ngadia; Frey, Jeremy; Schwartz, Martin; Bristiel, Philippe

2015-03-01

Analyzing crack stability in an industrial context is challenging due to the geometry of the structure. The finite element method is effective for defect-free problems. The boundary element method is effective for problems in simple geometries with singularities. We present a strategy that takes advantage of both approaches. Within the iterative solution procedure, the FEM solves a defect-free problem over the structure while the BEM solves the crack problem over a fictitious domain with simple geometry. The effectiveness of the approach is demonstrated on some simple examples which allow comparison with literature results and on an industrial problem.

Sieve tube geometry in relation to phloem flow.

PubMed

Mullendore, Daniel L; Windt, Carel W; Van As, Henk; Knoblauch, Michael

2010-03-01

Sieve elements are one of the least understood cell types in plants. Translocation velocities and volume flow to supply sinks with photoassimilates greatly depend on the geometry of the microfluidic sieve tube system and especially on the anatomy of sieve plates and sieve plate pores. Several models for phloem translocation have been developed, but appropriate data on the geometry of pores, plates, sieve elements, and flow parameters are lacking. We developed a method to clear cells from cytoplasmic constituents to image cell walls by scanning electron microscopy. This method allows high-resolution measurements of sieve element and sieve plate geometries. Sieve tube-specific conductivity and its reduction by callose deposition after injury was calculated for green bean (Phaseolus vulgaris), bamboo (Phyllostachys nuda), squash (Cucurbita maxima), castor bean (Ricinus communis), and tomato (Solanum lycopersicum). Phloem sap velocity measurements by magnetic resonance imaging velocimetry indicate that higher conductivity is not accompanied by a higher velocity. Studies on the temporal development of callose show that small sieve plate pores might be occluded by callose within minutes, but plants containing sieve tubes with large pores need additional mechanisms.

Sieve Tube Geometry in Relation to Phloem Flow

PubMed Central

Mullendore, Daniel L.; Windt, Carel W.; Van As, Henk; Knoblauch, Michael

2010-01-01

Sieve elements are one of the least understood cell types in plants. Translocation velocities and volume flow to supply sinks with photoassimilates greatly depend on the geometry of the microfluidic sieve tube system and especially on the anatomy of sieve plates and sieve plate pores. Several models for phloem translocation have been developed, but appropriate data on the geometry of pores, plates, sieve elements, and flow parameters are lacking. We developed a method to clear cells from cytoplasmic constituents to image cell walls by scanning electron microscopy. This method allows high-resolution measurements of sieve element and sieve plate geometries. Sieve tube–specific conductivity and its reduction by callose deposition after injury was calculated for green bean (Phaseolus vulgaris), bamboo (Phyllostachys nuda), squash (Cucurbita maxima), castor bean (Ricinus communis), and tomato (Solanum lycopersicum). Phloem sap velocity measurements by magnetic resonance imaging velocimetry indicate that higher conductivity is not accompanied by a higher velocity. Studies on the temporal development of callose show that small sieve plate pores might be occluded by callose within minutes, but plants containing sieve tubes with large pores need additional mechanisms. PMID:20354199

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Foote, F.G.; Jette, E.R.

1963-05-01

A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

Method and apparatus for diagnosing breached fuel elements

DOEpatents

Gross, K.C.; Lambert, J.D.B.; Nomura, S.

1987-03-02

The invention provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor. A detection system measures the activity of isotopes from the cover gas in the reactor. A data acquisition and processing system monitors the detection system and corrects for the effects of the cover-gas clean up system on the measured activity and further calculates the derivative curve of the corrected activity as a function of time. A plotting system graphs the derivative curve, which represents the instantaneous release rate of fission gas from a breached fuel element. 8 figs.

JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

DOEpatents

Szilard, L.; Wigner, E.P.; Creutz, E.C.

1959-05-12

Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

A machine learning approach as a surrogate of finite element analysis-based inverse method to estimate the zero-pressure geometry of human thoracic aorta.

PubMed

Liang, Liang; Liu, Minliang; Martin, Caitlin; Sun, Wei

2018-05-09

Advances in structural finite element analysis (FEA) and medical imaging have made it possible to investigate the in vivo biomechanics of human organs such as blood vessels, for which organ geometries at the zero-pressure level need to be recovered. Although FEA-based inverse methods are available for zero-pressure geometry estimation, these methods typically require iterative computation, which are time-consuming and may be not suitable for time-sensitive clinical applications. In this study, by using machine learning (ML) techniques, we developed an ML model to estimate the zero-pressure geometry of human thoracic aorta given 2 pressurized geometries of the same patient at 2 different blood pressure levels. For the ML model development, a FEA-based method was used to generate a dataset of aorta geometries of 3125 virtual patients. The ML model, which was trained and tested on the dataset, is capable of recovering zero-pressure geometries consistent with those generated by the FEA-based method. Thus, this study demonstrates the feasibility and great potential of using ML techniques as a fast surrogate of FEA-based inverse methods to recover zero-pressure geometries of human organs. Copyright © 2018 John Wiley & Sons, Ltd.

Focus in High School Mathematics: Reasoning and Sense Making in Geometry

ERIC Educational Resources Information Center

National Council of Teachers of Mathematics, 2010

2010-01-01

Classically, geometry has been the subject in which students encounter mathematical proof based on formal deduction. Attention to proof in the geometry curriculum is strengthened by a focus on reasoning and sense making. This book examines the four key elements (conjecturing about geometric objects, construction and evaluation of geometric…

The Epos of Euclidean Geometry in Greek Secondary Education (1836-1985): Pressure for Change and Resistance.

ERIC Educational Resources Information Center

Toumasis, Charalampos

1990-01-01

Presented is the development of school geometry in contemporary Greece since the era in which the secondary education system was established (1836). Identified and interpreted are the reasons for the endurance of Euclid's Elements as the basic component of school geometry in Greece. (KR)

Geometry optimization for micro-pressure sensor considering dynamic interference

NASA Astrophysics Data System (ADS)

Yu, Zhongliang; Zhao, Yulong; Li, Lili; Tian, Bian; Li, Cun

2014-09-01

Presented is the geometry optimization for piezoresistive absolute micro-pressure sensor. A figure of merit called the performance factor (PF) is defined as a quantitative index to describe the comprehensive performances of a sensor including sensitivity, resonant frequency, and acceleration interference. Three geometries are proposed through introducing islands and sensitive beams into typical flat diaphragm. The stress distributions of sensitive elements are analyzed by finite element method. Multivariate fittings based on ANSYS simulation results are performed to establish the equations about surface stress, deflection, and resonant frequency. Optimization by MATLAB is carried out to determine the dimensions of the geometries. Convex corner undercutting is evaluated. Each PF of the three geometries with the determined dimensions is calculated and compared. Silicon bulk micromachining is utilized to fabricate the prototypes of the sensors. The outputs of the sensors under both static and dynamic conditions are tested. Experimental results demonstrate the rationality of the defined performance factor and reveal that the geometry with quad islands presents the highest PF of 210.947 Hz1/4. The favorable overall performances enable the sensor more suitable for altimetry.

Multifuel evaluation of rich/quench/lean combustor

NASA Technical Reports Server (NTRS)

Notardonato, J. J.; Novick, A. S.; Troth, D. L.

1982-01-01

The fuel flexible combustor technology was developed for application to the Model 570-K industrial gas turbine engine. The technology, to achieve emission goals, emphasizes dry NOx reduction methods. Due to the high levels of fuel-bound nitrogen (FBN), control of NOx can be effected through a staged combustor with a rich initial combustion zone. A rich/quench/lean variable geometry combustor utilizes the technology presented to achieve low NOx from alternate fuels containing FBN. The results focus on emissions and durability for multifuel operation.

Gulf War Illnesses: DOD’s Conclusions about U.S. Troops’ Exposure Cannot be Adequately Supported

DTIC Science & Technology

2004-06-01

well fires, fumes from jet fuel , fumes from burning jet fuel in tents, petroleum in drinking water, depleted uranium munitions, smoking, alcohol use...Explosive 31 Figure 6: Boundary Layer Characteristics 32 Figure 7: Three Types of Plume Geometry 33 Figure 8: The Impact of Nocturnal Jets on a...ignited by thermite grenades—alone and with the addition of diesel fuel —as well as by fused initiation of the burster explosive charge. According to

Biomechanical analysis comparing natural and alloplastic temporomandibular joint replacement using a finite element model.

PubMed

Mesnard, Michel; Ramos, Antonio; Ballu, Alex; Morlier, Julien; Cid, M; Simoes, J A

2011-04-01

Prosthetic materials and bone present quite different mechanical properties. Consequently, mandible reconstruction with metallic materials (or a mandible condyle implant) modifies the physiologic behavior of the mandible (stress, strain patterns, and condyle displacements). The changing of bone strain distribution results in an adaptation of the temporomandibular joint, including articular contacts. Using a validated finite element model, the natural mandible strains and condyle displacements were evaluated. Modifications of strains and displacements were then assessed for 2 different temporomandibular joint implants. Because materials and geometry play important key roles, mechanical properties of cortical bone were taken into account in models used in finite element analysis. The finite element model allowed verification of the worst loading configuration of the mandibular condyle. Replacing the natural condyle by 1 of the 2 tested implants, the results also show the importance of the implant geometry concerning biomechanical mandibular behavior. The implant geometry and stiffness influenced mainly strain distribution. The different forces applied to the mandible by the elevator muscles, teeth, and joint loads indicate that the finite element model is a relevant tool to optimize implant geometry or, in a subsequent study, to choose a more suitable distribution of the screws. Bone screws (number and position) have a significant influence on mandibular behavior and on implant stress pattern. Stress concentration and implant fracture must be avoided. Copyright © 2011 American Association of Oral and Maxillofacial Surgeons. Published by Elsevier Inc. All rights reserved.

Enrichment Zoning Options for the Small Nuclear Rocket Engine (SNRE)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bruce G. Schnitzler; Stanley K. Borowski

2010-07-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. In NASA’s recent Mars Design Reference Architecture (DRA) 5.0 study (NASA-SP-2009-566, July 2009), nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option because of its high thrust and high specific impulse (-900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. An extensive nuclear thermal rocket technology development effortmore » was conducted from 1955-1973 under the Rover/NERVA Program. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art design incorporating lessons learned from the very successful technology development program. Past activities at the NASA Glenn Research Center have included development of highly detailed MCNP Monte Carlo transport models of the SNRE and other small engine designs. Preliminary core configurations typically employ fuel elements with fixed fuel composition and fissile material enrichment. Uniform fuel loadings result in undesirable radial power and temperature profiles in the engines. Engine performance can be improved by some combination of propellant flow control at the fuel element level and by varying the fuel composition. Enrichment zoning at the fuel element level with lower enrichments in the higher power elements at the core center and on the core periphery is particularly effective. Power flattening by enrichment zoning typically results in more uniform propellant exit temperatures and improved engine performance. For the SNRE, element enrichment zoning provided very flat radial power profiles with 551 of the 564 fuel elements within 1% of the average element power. Results for this and alternate enrichment zoning options for the SNRE are compared.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, Byoung Yoon; Roberts, Barry L.

The three-dimensional finite element mesh capturing realistic geometries of Bayou Choctaw site has been constructed using the sonar and seismic survey data obtained from the field. The mesh is consisting of hexahedral elements because the salt constitutive model is coded using hexahedral elements. Various ideas and techniques to construct finite element mesh capturing artificially and naturally formed geometries are provided. The techniques to reduce the number of elements as much as possible to save on computer run time with maintaining the computational accuracy is also introduced. The steps and methodologies could be applied to construct the meshes of Big Hill,more » Bryan Mound, and West Hackberry strategic petroleum reserve sites. The methodology could be applied to the complicated shape masses for not only various civil and geological structures but also biological applications such as artificial limbs.« less

Estimation of ICBM (Intercontinental Ballistic Missile) Performance Parameters

DTIC Science & Technology

1985-12-01

Formulation . . . . . 42 Staging Event Detection . . . . . . 43 Staging Estimator for Two State System . 46 * Staging Time and Vehicle Parameter...6 4. Land Based Sensor Coordinate System . . . . 10 5. Radar Site Geometry . . . . . . . . . 1 6. Observation Geometry . . . . . . . . . 12 7...Ve dm mi + dmi+d Figure 3. Rocket Thrust of fuel, the equation of motion of the rocket can be devel--S oped. This is a closed system of particles

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

DOEpatents

Brooks, H.

1960-04-26

A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

Ion chromatographic determination of sulfur in fuels

NASA Technical Reports Server (NTRS)

Mizisin, C. S.; Kuivinen, D. E.; Otterson, D. A.

1978-01-01

The sulfur content of fuels was determined using an ion chromatograph to measure the sulfate produced by a modified Parr bomb oxidation. Standard Reference Materials from the National Bureau of Standards, of approximately 0.2 + or - 0.004% sulfur, were analyzed resulting in a standard deviation no greater than 0.008. The ion chromatographic method can be applied to conventional fuels as well as shale-oil derived fuels. Other acid forming elements, such as fluorine, chlorine and nitrogen could be determined at the same time, provided that these elements have reached a suitable ionic state during the oxidation of the fuel.

DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Hofman, G.L.

1999-04-01

Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al in variously designed dispersion fuel elements.

X-ray Absorption Spectroscopy Characterization of Electrochemical Processes in Renewable Energy Storage and Conversion Devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmand, Maryam

2013-05-19

The development of better energy conversion and storage devices, such as fuel cells and batteries, is crucial for reduction of our global carbon footprint and improving the quality of the air we breathe. However, both of these technologies face important challenges. The development of lower cost and better electrode materials, which are more durable and allow more control over the electrochemical reactions occurring at the electrode/electrolyte interface, is perhaps most important for meeting these challenges. Hence, full characterization of the electrochemical processes that occur at the electrodes is vital for intelligent design of more energy efficient electrodes. X-ray absorption spectroscopymore » (XAS) is a short-range order, element specific technique that can be utilized to probe the processes occurring at operating electrode surfaces, as well for studying the amorphous materials and nano-particles making up the electrodes. It has been increasingly used in recent years to study fuel cell catalysts through application of the and #916; and mgr; XANES technique, in combination with the more traditional X-ray Absorption Near Edge Structure (XANES) and Extended X-ray Absorption Fine Structure (EXAFS) techniques. The and #916; and mgr; XANES data analysis technique, previously developed and applied to heterogeneous catalysts and fuel cell electrocatalysts by the GWU group, was extended in this work to provide for the first time space resolved adsorbate coverages on both electrodes of a direct methanol fuel cell. Even more importantly, the and #916; and mgr; technique was applied for the first time to battery relevant materials, where bulk properties such as the oxidation state and local geometry of a cathode are followed.« less

A three-dimensional finite element model of human atrial anatomy: New methods for cubic Hermite meshes with extraordinary vertices

PubMed Central

Gonzales, Matthew J.; Sturgeon, Gregory; Krishnamurthy, Adarsh; Hake, Johan; Jonas, René; Stark, Paul; Rappel, Wouter-Jan; Narayan, Sanjiv M.; Zhang, Yongjie; Segars, W. Paul; McCulloch, Andrew D.

2013-01-01

High-order cubic Hermite finite elements have been valuable in modeling cardiac geometry, fiber orientations, biomechanics, and electrophysiology, but their use in solving three-dimensional problems has been limited to ventricular models with simple topologies. Here, we utilized a subdivision surface scheme and derived a generalization of the “local-to-global” derivative mapping scheme of cubic Hermite finite elements to construct bicubic and tricubic Hermite models of the human atria with extraordinary vertices from computed tomography images of a patient with atrial fibrillation. To an accuracy of 0.6 millimeters, we were able to capture the left atrial geometry with only 142 bicubic Hermite finite elements, and the right atrial geometry with only 90. The left and right atrial bicubic Hermite meshes were G1 continuous everywhere except in the one-neighborhood of extraordinary vertices, where the mean dot products of normals at adjacent elements were 0.928 and 0.925. We also constructed two biatrial tricubic Hermite models and defined fiber orientation fields in agreement with diagrammatic data from the literature using only 42 angle parameters. The meshes all have good quality metrics, uniform element sizes, and elements with aspect ratios near unity, and are shared with the public. These new methods will allow for more compact and efficient patient-specific models of human atrial and whole heart physiology. PMID:23602918

Performance Analysis and Modeling of Thermally Sprayed Resistive Heaters

NASA Astrophysics Data System (ADS)

Lamarre, Jean-Michel; Marcoux, Pierre; Perrault, Michel; Abbott, Richard C.; Legoux, Jean-Gabriel

2013-08-01

Many processes and systems require hot surfaces. These are usually heated using electrical elements located in their vicinity. However, this solution is subject to intrinsic limitations associated with heating element geometry and physical location. Thermally spraying electrical elements directly on surfaces can overcome these limitations by tailoring the geometry of the heating element to the application. Moreover, the element heat transfer is maximized by minimizing the distance between the heater and the surface to be heated. This article is aimed at modeling and characterizing resistive heaters sprayed on metallic substrates. Heaters were fabricated by using a plasma-sprayed alumina dielectric insulator and a wire flame-sprayed iron-based alloy resistive element. Samples were energized and kept at a constant temperature of 425 °C for up to 4 months. SEM cross-sectional observations revealed the formation of cracks at very specific locations in the alumina layer after thermal use. Finite-element modeling shows that these cracks originate from high local thermal stresses and can be predicted according to the considered geometry. The simulation model was refined using experimental parameters obtained by several techniques such as emissivity and time-dependent temperature profile (infra-red camera), resistivity (four-probe technique), thermal diffusivity (laser flash method), and mechanical properties (micro and nanoindentation). The influence of the alumina thickness and the substrate material on crack formation was evaluated.

The non-linear response of a muscle in transverse compression: assessment of geometry influence using a finite element model.

PubMed

Gras, Laure-Lise; Mitton, David; Crevier-Denoix, Nathalie; Laporte, Sébastien

2012-01-01

Most recent finite element models that represent muscles are generic or subject-specific models that use complex, constitutive laws. Identification of the parameters of such complex, constitutive laws could be an important limit for subject-specific approaches. The aim of this study was to assess the possibility of modelling muscle behaviour in compression with a parametric model and a simple, constitutive law. A quasi-static compression test was performed on the muscles of dogs. A parametric finite element model was designed using a linear, elastic, constitutive law. A multi-variate analysis was performed to assess the effects of geometry on muscle response. An inverse method was used to define Young's modulus. The non-linear response of the muscles was obtained using a subject-specific geometry and a linear elastic law. Thus, a simple muscle model can be used to have a bio-faithful, biomechanical response.

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searchedmore » and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.« less

Simplified jet fuel reaction mechanism for lean burn combustion application

NASA Technical Reports Server (NTRS)

Lee, Chi-Ming; Kundu, Krishna; Ghorashi, Bahman

1993-01-01

Successful modeling of combustion and emissions in gas turbine engine combustors requires an adequate description of the reaction mechanism. Detailed mechanisms contain a large number of chemical species participating simultaneously in many elementary kinetic steps. Current computational fluid dynamic models must include fuel vaporization, fuel-air mixing, chemical reactions, and complicated boundary geometries. A five-step Jet-A fuel mechanism which involves pyrolysis and subsequent oxidation of paraffin and aromatic compounds is presented. This mechanism is verified by comparing with Jet-A fuel ignition delay time experimental data, and species concentrations obtained from flametube experiments. This five-step mechanism appears to be better than the current one- and two-step mechanisms.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Methods for making a porous nuclear fuel element

DOEpatents

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Remarks on the foundations of geometry and immersion theory

NASA Astrophysics Data System (ADS)

Odon, P. I.; Capistrano, A. J. S.

2010-04-01

In this paper, we deal with the evolution of physics and maths, and how one is intrinsically connected to the other. Euclid and his book Elements, and the importance of the fifth postulate for geometry led to the discovery of non-Euclidean geometries. We point out how these geometries play an essential role in immersion theory and Nash's theorem, and its importance for physics when applied to the brane-world theory.

Upgraded HFIR Fuel Element Welding System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sease, John D

2010-02-01

The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. Inmore » recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.« less

Optimizing detector geometry for trace element mapping by X-ray fluorescence.

PubMed

Sun, Yue; Gleber, Sophie-Charlotte; Jacobsen, Chris; Kirz, Janos; Vogt, Stefan

2015-05-01

Trace metals play critical roles in a variety of systems, ranging from cells to photovoltaics. X-Ray Fluorescence (XRF) microscopy using X-ray excitation provides one of the highest sensitivities available for imaging the distribution of trace metals at sub-100 nm resolution. With the growing availability and increasing performance of synchrotron light source based instruments and X-ray nanofocusing optics, and with improvements in energy-dispersive XRF detectors, what are the factors that limit trace element detectability? To address this question, we describe an analytical model for the total signal incident on XRF detectors with various geometries, including the spectral response of energy dispersive detectors. This model agrees well with experimentally recorded X-ray fluorescence spectra, and involves much shorter calculation times than with Monte Carlo simulations. With such a model, one can estimate the signal when a trace element is illuminated with an X-ray beam, and when just the surrounding non-fluorescent material is illuminated. From this signal difference, a contrast parameter can be calculated and this can in turn be used to calculate the signal-to-noise ratio (S/N) for detecting a certain elemental concentration. We apply this model to the detection of trace amounts of zinc in biological materials, and to the detection of small quantities of arsenic in semiconductors. We conclude that increased detector collection solid angle is (nearly) always advantageous even when considering the scattered signal. However, given the choice between a smaller detector at 90° to the beam versus a larger detector at 180° (in a backscatter-like geometry), the 90° detector is better for trace element detection in thick samples, while the larger detector in 180° geometry is better suited to trace element detection in thin samples. Copyright © 2015. Published by Elsevier B.V.

Optimizing detector geometry for trace element mapping by X-ray fluorescence

PubMed Central

Sun, Yue; Gleber, Sophie-Charlotte; Jacobsen, Chris; Kirz, Janos; Vogt, Stefan

2016-01-01

Trace metals play critical roles in a variety of systems, ranging from cells to photovoltaics. X-Ray Fluorescence (XRF) microscopy using X-ray excitation provides one of the highest sensitivities available for imaging the distribution of trace metals at sub-100 nm resolution. With the growing availability and increasing performance of synchrotron light source based instruments and X-ray nanofocusing optics, and with improvements in energy-dispersive XRF detectors, what are the factors that limit trace element detectability? To address this question, we describe an analytical model for the total signal incident on XRF detectors with various geometries, including the spectral response of energy dispersive detectors. This model agrees well with experimentally recorded X-ray fluorescence spectra, and involves much shorter calculation times than with Monte Carlo simulations. With such a model, one can estimate the signal when a trace element is illuminated with an X-ray beam, and when just the surrounding non-fluorescent material is illuminated. From this signal difference, a contrast parameter can be calculated and this can in turn be used to calculate the signal-to-noise ratio (S/N) for detecting a certain elemental concentration. We apply this model to the detection of trace amounts of zinc in biological materials, and to the detection of small quantities of arsenic in semiconductors. We conclude that increased detector collection solid angle is (nearly) always advantageous even when considering the scattered signal. However, given the choice between a smaller detector at 90° to the beam versus a larger detector at 180° (in a backscatter-like geometry), the 90° detector is better for trace element detection in thick samples, while the larger detector in 180° geometry is better suited to trace element detection in thin samples. PMID:25600825

Optimizing detector geometry for trace element mapping by X-ray fluorescence

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Yue; Gleber, Sophie-Charlotte; Jacobsen, Chris

Trace metals play critical roles in a variety of systems, ranging from cells to photovoltaics. X-Ray Fluorescence (XRF) microscopy using X-ray excitation provides one of the highest sensitivities available for imaging the distribution of trace metals at sub-100 nm resolution. With the growing availability and increasing performance of synchrotron light source based instruments and X-ray nanofocusing optics, and with improvements in energy-dispersive XRF detectors, what are the factors that limit trace element detectability? To address this question, we describe an analytical model for the total signal incident on XRF detectors with various geometries, including the spectral response of energy dispersivemore » detectors. This model agrees well with experimentally recorded X-ray fluorescence spectra, and involves much shorter calculation times than with Monte Carlo simulations. With such a model, one can estimate the signal when a trace element is illuminated with an X-ray beam, and when just the surrounding non-fluorescent material is illuminated. From this signal difference, a contrast parameter can be calculated and this can in turn be used to calculate the signal-to-noise ratio (S/N) for detecting a certain elemental concentration. We apply this model to the detection of trace amounts of zinc in biological materials, and to the detection of small quantities of arsenic in semiconductors. We conclude that increased detector collection solid angle is (nearly) always advantageous even when considering the scattered signal. However, given the choice between a smaller detector at 90° to the beam versus a larger detector at 180° (in a backscatter-like geometry), the 90° detector is better for trace element detection in thick samples, while the larger detector in 180° geometry is better suited to trace element detection in thin samples.« less

Optical frequency selective surface design using a GPU accelerated finite element boundary integral method

NASA Astrophysics Data System (ADS)

Ashbach, Jason A.

Periodic metallodielectric frequency selective surface (FSS) designs have historically seen widespread use in the microwave and radio frequency spectra. By scaling the dimensions of an FSS unit cell for use in a nano-fabrication process, these concepts have recently been adapted for use in optical applications as well. While early optical designs have been limited to wellunderstood geometries or optimized pixelated screens, nano-fabrication, lithographic and interconnect technology has progressed to a point where it is possible to fabricate metallic screens of arbitrary geometries featuring curvilinear or even three-dimensional characteristics that are only tens of nanometers wide. In order to design an FSS featuring such characteristics, it is important to have a robust numerical solver that features triangular elements in purely two-dimensional geometries and prismatic or tetrahedral elements in three-dimensional geometries. In this dissertation, a periodic finite element method code has been developed which features prismatic elements whose top and bottom boundaries are truncated by numerical integration of the boundary integral as opposed to an approximate representation found in a perfectly matched layer. However, since no exact solution exists for the calculation of triangular elements in a boundary integral, this process can be time consuming. To address this, these calculations were optimized for parallelization such that they may be done on a graphics processor, which provides a large increase in computational speed. Additionally, a simple geometrical representation using a Bezier surface is presented which provides generality with few variables. With a fast numerical solver coupled with a lowvariable geometric representation, a heuristic optimization algorithm has been used to develop several optical designs such as an absorber, a circular polarization filter, a transparent conductive surface and an enhanced, optical modulator.

A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

2007-12-01

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

Apollo 12 Mission image - Alan Bean unloads ALSEP RTG fuel element

NASA Image and Video Library

1969-11-19

AS12-46-6790 (19 Nov. 1969) --- Astronaut Alan L. Bean, lunar module pilot, is photographed at quadrant II of the Lunar Module (LM) during the first Apollo 12 extravehicular activity (EVA) on the moon. This picture was taken by astronaut Charles Conrad Jr., commander. Here, Bean is using a fuel transfer tool to remove the fuel element from the fuel cask mounted on the LM's descent stage. The fuel element was then placed in the Radioisotope Thermoelectric Generator (RTG), the power source for the Apollo Lunar Surface Experiments Package (ALSEP) which was deployed on the moon by the two astronauts. The RTG is next to Bean's right leg. While astronauts Conrad and Bean descended in the LM "Intrepid" to explore the Ocean of Storms region of the moon, astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) "Yankee Clipper" in lunar orbit.

Reliability analysis of dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

2008-03-01

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Method for recovering catalytic elements from fuel cell membrane electrode assemblies

DOEpatents

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

2012-06-26

A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

Combined catalysts for the combustion of fuel in gas turbines

DOEpatents

Anoshkina, Elvira V.; Laster, Walter R.

2012-11-13

A catalytic oxidation module for a catalytic combustor of a gas turbine engine is provided. The catalytic oxidation module comprises a plurality of spaced apart catalytic elements for receiving a fuel-air mixture over a surface of the catalytic elements. The plurality of catalytic elements includes at least one primary catalytic element comprising a monometallic catalyst and secondary catalytic elements adjacent the primary catalytic element comprising a multi-component catalyst. Ignition of the monometallic catalyst of the primary catalytic element is effective to rapidly increase a temperature within the catalytic oxidation module to a degree sufficient to ignite the multi-component catalyst.

Development of procedures for calculating stiffness and damping properties of elastomers. Part 3: The effects of temperature, dissipation level and geometry

NASA Technical Reports Server (NTRS)

Smalley, A. J.; Tessarzik, J. M.

1975-01-01

Effects of temperature, dissipation level and geometry on the dynamic behavior of elastomer elements were investigated. Force displacement relationships in elastomer elements and the effects of frequency, geometry and temperature upon these relationships are reviewed. Based on this review, methods of reducing stiffness and damping data for shear and compression test elements to material properties (storage and loss moduli) and empirical geometric factors are developed and tested using previously generated experimental data. A prediction method which accounts for large amplitudes of deformation is developed on the assumption that their effect is to increase temperature through the elastomers, thereby modifying the local material properties. Various simple methods of predicting the radial stiffness of ring cartridge elements are developed and compared. Material properties were determined from the shear specimen tests as a function of frequency and temperature. Using these material properties, numerical predictions of stiffness and damping for cartridge and compression specimens were made and compared with corresponding measurements at different temperatures, with encouraging results.

The geometry of the forisome-sieve element-sieve plate complex in the phloem of Vicia faba L. leaflets.

PubMed

Peters, Winfried S; van Bel, Aart J E; Knoblauch, Michael

2006-01-01

Forisomes are contractile protein bodies that appear to control flux rates in the phloem of faboid legumes by reversibly plugging the sieve tubes. Plugging is triggered by Ca(2+) which induces an anisotropic deformation of forisomes, consisting of a longitudinal contraction and a radial expansion. By conventional light microscopy and confocal laser-scanning microscopy, the three-dimensional geometry of the forisome-sieve element-sieve plate complex in intact sieve tubes of leaflets of Vicia faba L. was reconstructed. Forisomes were mostly located close to sieve plates, and occasionally were observed drifting unrestrainedly along the sieve element, suggesting that they might be utilized as internal markers of flow direction. The diameter of forisomes in the resting state correlated with the diameter of their sieve elements, supporting the idea that radial expansion of forisomes is the geometric basis of reversible sieve tube plugging. Comparison of the present results regarding forisome geometry in situ with previously published data on forisome reactivity in vitro makes it questionable, however, whether forisomes are capable of completely sealing sieve tubes in V. faba leaves.

Finite element techniques applied to cracks interacting with selected singularities

NASA Technical Reports Server (NTRS)

Conway, J. C.

1975-01-01

The finite-element method for computing the extensional stress-intensity factor for cracks approaching selected singularities of varied geometry is described. Stress-intensity factors are generated using both displacement and J-integral techniques, and numerical results are compared to those obtained experimentally in a photoelastic investigation. The selected singularities considered are a colinear crack, a circular penetration, and a notched circular penetration. Results indicate that singularities greatly influence the crack-tip stress-intensity factor as the crack approaches the singularity. In addition, the degree of influence can be regulated by varying the overall geometry of the singularity. Local changes in singularity geometry have little effect on the stress-intensity factor for the cases investigated.

THE FUEL ELEMENT GRAPHITE. Project DRAGON.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Graham, L.W.; Price, M.S.T.

1963-01-15

The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

Low exchange element for nuclear reactor

DOEpatents

Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

1985-01-01

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

Emission response from extended length, variable geometry gas turbine combustor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Troth, D.L.; Verdouw, A.J.; Tomlinson, J.G.

1974-01-01

A program to analyze, select, and experimentally evaluate low emission combustors for aircraft gas turbine engines is conducted to demonstrate a final combustor concept having a 50 percent reduction in total mass emissions (carbon monoxide, unburnt hydrocarbons, oxides of nitrogen, and exhaust smoke) without an increase in any specific pollutant. Research conducted under an Army Contract established design concepts demonstrating significant reductions in CO and UHC emissions. Two of these concepts were an extended length intermediate zone to consume CO and UHC and variable geometry to control the primary zone fuel air ratio over varying power conditions. Emission reduction featuresmore » were identified by analytical methods employing both reaction kinetics and empirical correlations. Experimental results were obtained on a T63 component combustor rig operating at conditions simulating the engine over the complete power operating range with JP-4 fuel. A combustor incorporating both extended length and variable geometry was evaluated and the performance and emission results are reported. These results are compared on the basis of a helicopter duty cycle and the EPA 1979 turboprop regulation landing take off cycle. The 1979 EPA emission regulations for P2 class engines can be met with the extended length variable geometry combustor on the T63 turboprop engine.« less

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit corrosion of certain potential NTR fuel forms. Additional diagnostic upgrades included in the present NTREES set up include the addition of a gamma ray spectrometer located near the vent filter to detect uranium fuel particles exiting the fuel element in the propellant exhaust stream to provide additional information any material loss occurring during testing. Other aspects of the upgrade included reworking NTREES to reduce the operational complexity of the system despite the increased complexity of the induction heating system. To this end, many of the controls were consolidated on fewer panels. As part of this upgrade activity, the Safety Assessment (SA) and the Standard Operating Procedures (SOPs) for NTREES were extensively rewritten. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can be accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements.

Modeling Crustal Deformation Due to the Landers, Hector Mine Earthquakes Using the SCEC Community Fault Model

NASA Astrophysics Data System (ADS)

Gable, C. W.; Fialko, Y.; Hager, B. H.; Plesch, A.; Williams, C. A.

2006-12-01

More realistic models of crustal deformation are possible due to advances in measurements and modeling capabilities. This study integrates various data to constrain a finite element model of stress and strain in the vicinity of the 1992 Landers earthquake and the 1999 Hector Mine earthquake. The geometry of the model is designed to incorporate the Southern California Earthquake Center (SCEC), Community Fault Model (CFM) to define fault geometry. The Hector Mine fault is represented by a single surface that follows the trace of the Hector Mine fault, is vertical and has variable depth. The fault associated with the Landers earthquake is a set of seven surfaces that capture the geometry of the splays and echelon offsets of the fault. A three dimensional finite element mesh of tetrahedral elements is built that closely maintains the geometry of these fault surfaces. The spatially variable coseismic slip on faults is prescribed based on an inversion of geodetic (Synthetic Aperture Radar and Global Positioning System) data. Time integration of stress and strain is modeled with the finite element code Pylith. As a first step the methodology of incorporating all these data is described. Results of the time history of the stress and strain transfer between 1992 and 1999 are analyzed as well as the time history of deformation from 1999 to the present.

NUCLEAR REACTOR CORE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, F.R.

1963-02-01

A nuclear reactor core composed of a number of identical elements of solid moderator material fitted together was designed. Each moderator element is apertured to provide channels for fuel and coolant. The elements have an external shape which permits them to be stacked in layers with similar elements, with the surfaces of adjacent elements fitting and in contact with each other. The cross section of the element is of a general hexagonal shape with identations and protrusions, so that the elements can be fitted together. The described core should not be liable to fracture under transverse loading. Specific arrangements ofmore » moderator elements and fuel and coolant apertures are described. (M.P.G.)« less

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.

As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of themore » cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.« less

Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

PubMed

Saqib, Naeem; Bäckström, Mattias

2014-12-01

Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature. Copyright © 2014 Elsevier Ltd. All rights reserved.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-01

A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

NASA Broad Specification Fuels Combustion Technology program - Pratt and Whitney Aircraft Phase I results and status

NASA Technical Reports Server (NTRS)

Lohmann, R. P.; Fear, J. S.

1982-01-01

In connection with increases in the cost of fuels and the reduced availability of high quality petroleum crude, a modification of fuel specifications has been considered to allow acceptance of poorer quality fuels. To obtain the information upon which a selection of appropriate fuels for aircraft can be based, the Broad Specification Fuels Combustion Technology program was formulated by NASA. A description is presented of program-related investigations conducted by an American aerospace company. The specific objective of Phase I of this program has been to evaluate the impact of the use of broadened properties fuels on combustor design through comprehensive combustor rig testing. Attention is given to combustor concepts, experimental evaluation, results obtained with single stage combustors, the stage combustor concept, and the capability of a variable geometry combustor.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes thses overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rowsell, David Leon

This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

Demonstration of Subscale Cermet Fuel Specimen Fabrication Approach Using Spark Plasma Sintering and Diffusion Bonding

NASA Technical Reports Server (NTRS)

Barnes, Marvin W.; Tucker, Dennis S.; Benensky, Kelsa M.

2018-01-01

Nuclear thermal propulsion (NTP) has the potential to expand the limits of human space exploration by enabling crewed missions to Mars and beyond. The viability of NTP hinges on the development of a robust nuclear fuel material that can perform in the harsh operating environment (> or = 2500K, reactive hydrogen) of a nuclear thermal rocket (NTR) engine. Efforts are ongoing to develop fuel material and to assemble fuel elements that will be stable during the service life of an NTR. Ceramic-metal (cermet) fuels are being actively pursued by NASA Marshall Space Flight Center (MSFC) due to their demonstrated high-temperature stability and hydrogen compatibility. Building on past cermet fuel development research, experiments were conducted to investigate a modern fabrication approach for cermet fuel elements. The experiments used consolidated tungsten (W)-60vol%zirconia (ZrO2) compacts that were formed via spark plasma sintering (SPS). The consolidated compacts were stacked and diffusion bonded to assess the integrity of the bond lines and internal cooling channel cladding. The assessment included hot hydrogen testing of the manufactured surrogate fuel and pure W for 45 minutes at 2500 K in the compact fuel element environmental test (CFEET) system. Performance of bonded W-ZrO2 rods was compared to bonded pure W rods to access bond line integrity and composite stability. Bonded surrogate fuels retained structural integrity throughout testing and incurred minimal mass loss.

The manufacture of LEU fuel elements at Dounreay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gibson, J.

1997-08-01

Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

Corrosion protected, multi-layer fuel cell interface

DOEpatents

Feigenbaum, Haim; Pudick, Sheldon; Wang, Chiu L.

1986-01-01

An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. The multi-layer configuration for the interface comprises a non-cupreous metal-coated metallic element to which is film-bonded a conductive layer by hot pressing a resin therebetween. The multi-layer arrangement provides bridging electrical contact.

A numerical investigation of the influence of radiation and moisture content on pyrolysis and ignition of a leaf-like fuel element

Treesearch

B.L. Yashwanth; B. Shotorban; S. Mahalingam; C.W. Lautenberger; David Weise

2016-01-01

The effects of thermal radiation and moisture content on the pyrolysis and gas phase ignition of a solid fuel element containing high moisture content were investigated using the coupled Gpyro3D/FDS models. The solid fuel has dimensions of a typical Arctostaphylos glandulosa leaf which is modeled as thin cellulose subjected to radiative heating on...

AST Critical Propulsion and Noise Reduction Technologies for Future Commercial Subsonic Engines Area of Interest 1.0: Reliable and Affordable Control Systems

NASA Technical Reports Server (NTRS)

Myers, William; Winter, Steve

2006-01-01

The General Electric Reliable and Affordable Controls effort under the NASA Advanced Subsonic Technology (AST) Program has designed, fabricated, and tested advanced controls hardware and software to reduce emissions and improve engine safety and reliability. The original effort consisted of four elements: 1) a Hydraulic Multiplexer; 2) Active Combustor Control; 3) a Variable Displacement Vane Pump (VDVP); and 4) Intelligent Engine Control. The VDVP and Intelligent Engine Control elements were cancelled due to funding constraints and are reported here only to the state they progressed. The Hydraulic Multiplexing element developed and tested a prototype which improves reliability by combining the functionality of up to 16 solenoids and servo-valves into one component with a single electrically powered force motor. The Active Combustor Control element developed intelligent staging and control strategies for low emission combustors. This included development and tests of a Controlled Pressure Fuel Nozzle for fuel sequencing, a Fuel Multiplexer for individual fuel cup metering, and model-based control logic. Both the Hydraulic Multiplexer and Controlled Pressure Fuel Nozzle system were cleared for engine test. The Fuel Multiplexer was cleared for combustor rig test which must be followed by an engine test to achieve full maturation.

Modelling of diesel engine fuelled with biodiesel using engine simulation software

NASA Astrophysics Data System (ADS)

Said, Mohd Farid Muhamad; Said, Mazlan; Aziz, Azhar Abdul

2012-06-01

This paper is about modelling of a diesel engine that operates using biodiesel fuels. The model is used to simulate or predict the performance and combustion of the engine by simplified the geometry of engine component in the software. The model is produced using one-dimensional (1D) engine simulation software called GT-Power. The fuel properties library in the software is expanded to include palm oil based biodiesel fuels. Experimental works are performed to investigate the effect of biodiesel fuels on the heat release profiles and the engine performance curves. The model is validated with experimental data and good agreement is observed. The simulation results show that combustion characteristics and engine performances differ when biodiesel fuels are used instead of no. 2 diesel fuel.

ERBS fuel addendum: Pollution reduction technology program small jet aircraft engines, phase 3

NASA Technical Reports Server (NTRS)

Bruce, T. W.; Davis, F. G.; Kuhn, T. E.; Mongia, H. C.

1982-01-01

A Model TFE731-2 engine with a low emission, variable geometry combustion system was tested to compare the effects of operating the engine on Commercial Jet-A aviation turbine fuel and experimental referee broad specification (ERBS) fuels. Low power emission levels were essentially identical while the high power NOx emission indexes were approximately 15% lower with the EBRS fuel. The exhaust smoke number was approximately 50% higher with ERBS at the takeoff thrust setting; however, both values were still below the EPA limit of 40 for the Model TFE731 engine. Primary zone liner wall temperature ran an average of 25 K higher with ERBS fuel than with Jet-A. The possible adoption of broadened proprties fuels for gas turbine applications is suggested.

Heat exchangers in regenerative gas turbine cycles

NASA Astrophysics Data System (ADS)

Nina, M. N. R.; Aguas, M. P. N.

1985-09-01

Advances in compact heat exchanger design and fabrication together with fuel cost rises continuously improve the attractability of regenerative gas turbine helicopter engines. In this study cycle parameters aiming at reduced specific fuel consumption and increased payload or mission range, have been optimized together with heat exchanger type and size. The discussion is based on a typical mission for an attack helicopter in the 900 kw power class. A range of heat exchangers is studied to define the most favorable geometry in terms of lower fuel consumption and minimum engine plus fuel weight. Heat exchanger volume, frontal area ratio and pressure drop effect on cycle efficiency are considered.

Experimental Extinguishment of Fires by Blast.

DTIC Science & Technology

1982-05-01

Icopmy DTIC TAB I Unan-nounced Approved by: Justiri cat i O__ : R. C. Phillips, Director Chemical Engineering Laboratory D1str1i.:: to,/ G. R...SRI study. Of interest this year has been the role played by fuel type--notably, the Dertinent physico- chemical properties--compared with the...perturbing geometries on fire behavior. Present tests on common liquid fuels representing various combinations of physico- chemical properties demonstrate

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

DOE PAGES

David, Stan A.; Miller, Roger G.; Feng, Zhili

2016-08-31

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

David, Stan A.; Miller, Roger G.; Feng, Zhili

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

The Nuclear Cryogenic Propulsion Stage

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Broadway, Jeramie W.; Gerrish, Harold P.; Belvin, Anthony D.; Borowski, Stanley K.; Scott, John H.

2014-01-01

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

SPOC Benchmark Case: SNRE Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vishal Patel; Michael Eades; Claude Russel Joyner II

The Small Nuclear Rocket Engine (SNRE) was modeled in the Center for Space Nuclear Research’s (CSNR) Space Propulsion Optimization Code (SPOC). SPOC aims to create nuclear thermal propulsion (NTP) geometries quickly to perform parametric studies on design spaces of historic and new NTP designs. The SNRE geometry was modeled in SPOC and a critical core with a reasonable amount of criticality margin was found. The fuel, tie-tubes, reflector, and control drum masses were predicted rather well. These are all very important for neutronics calculations so the active reactor geometries created with SPOC can continue to be trusted. Thermal calculations ofmore » the average and hot fuel channels agreed very well. The specific impulse calculations used historically and in SPOC disagree so mass flow rates and impulses differed. Modeling peripheral and power balance components that do not affect nuclear characteristics of the core is not a feature of SPOC and as such, these components should continue to be designed using other tools. A full paper detailing the available SNRE data and comparisons with SPOC outputs will be submitted as a follow-up to this abstract.« less

An experimental investigation of rubbing interaction in labyrinth seals at cryogenic temperature

NASA Technical Reports Server (NTRS)

Dolan, F. X.; Kennedy, F. E.; Schulson, E. M.

1985-01-01

An experimental program was carried out to address issues related to the observed cracking of the titanium knife edges on the labyrinth seals of the high pressure fuel pump (HPFP) in the Space Shuttle main engine (SSME). Thermal shock experiments were carried out using a jet specimen with geometry similar to the knife edge geometry. These tests demonstrate that cracking of the titanium alloy is possible in a situation involving repeated thermal cycles over a wide temperature range, as might be realized during a rub in the liquid hydrogen fuel pump. High speed rub interaction tests were conducted using a representative knife edge and seal geometry over a broad range of interaction rates. Alternative materials were also experimentally evaluated. These tests provide information which can be used to design improved labyrinth seals for the HPFP of the SSME. In particular, plasma-sprayed aluminum-graphite was found to be significantly better than aluminum alloy seals used at present from the standpoint of rub performance. Ion nitriding of the titanium alloy knife edges was also found to improve rub performance compared with the untreated baseline knife edge material.

Component testing of a ground based gas turbine steam cooled rich-burn primary zone combustor for emissions control of nitrogeneous fuels

NASA Technical Reports Server (NTRS)

Schultz, D. F.

1986-01-01

This effort summarizes the work performed on a steam cooled, rich-burn primary zone, variable geometry combustor designed for combustion of nitrogeneous fuels such as heavy oils or synthetic crude oils. The steam cooling was employed to determine its feasibility and assess its usefulness as part of a ground based gas turbine bottoming cycle. Variable combustor geometry was employed to demonstrate its ability to control primary and secondary zone equivalence ratios and overall pressure drop. Both concepts proved to be highly successful in achieving their desired objectives. The steam cooling reduced peak liner temperatures to less than 800 K. This low temperature offers the potential of both long life and reduced use of strategic materials for liner fabrication. These degrees of variable geometry were successfully employed to control air flow distribution within the combustor. A variable blade angle axial flow air swirler was used to control primary zone air flow, while the secondary and tertiary zone air flows were controlled by rotating bands which regulated air flow to the secondary zone quench holes and the dilutions holes respectively.

Development, Testing and Validation of a Waste Assay System for the Measurement and Characterisation of Active Spent Fuel Element Debris From UK Magnox Reactors - 12533

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mason, John A.; Burke, Kevin J.; Looman, Marc R.

2012-07-01

This paper describes the development, testing and validation of a waste measurement instrument for characterising active remote handled radioactive waste arising from the operation of Magnox reactors in the United Kingdom. Following operation in UK Magnox gas cooled reactors and a subsequent period of cooling, parts of the magnesium-aluminium alloy cladding were removed from spent fuel and the uranium fuel rods with the remaining cladding were removed to Sellafield for treatment. The resultant Magnox based spent fuel element debris (FED), which constitutes active intermediate level waste (ILW) has been stored in concrete vaults at the reactor sites. As part ofmore » the decommissioning of the FED vaults the FED must be removed, measured and characterised and placed in intermediate storage containers. The present system was developed for use at the Trawsfynydd nuclear power station (NPS), which is in the decommissioning phase, but the approach is potentially applicable to FED characterisation at all of the Magnox reactors. The measurement system consists of a heavily shielded and collimated high purity Germanium (HPGe) detector with electromechanical cooling and a high count-rate preamplifier and digital multichannel pulse height analyser. The HPGe based detector system is controlled by a software code, which stores the measurement result and allows a comprehensive analysis of the measured FED data. Fuel element debris is removed from the vault and placed on a tray to a uniform depth of typically 10 cm for measurement. The tray is positioned approximately 1.2 meters above the detector which views the FED through a tungsten collimator with an inverted pyramid shape. At other Magnox sites the positions may be reversed with the shielded and collimated HPGe detector located above the tray on which the FED is measured. A comprehensive Monte Carlo modelling and analysis of the measurement process has been performed in order to optimise the measurement geometry and eliminate interferences from radioactive sources and FED in the immediate vicinity of the measurement position. The detector system has been calibrated and high activity radioactive sources of Cs-137, Co-60 and Na-22 have been used to validate the measurement process. The data acquisition and analysis software code has been tested and validated in keeping with the software quality assurance requirements of both ISO:9001-2008 - TICK-IT in the UK and NQA-1. The measurement and analysis system has been comprehensively tested with high activity sources, is flexible and may be applicable to a wide range of remote handled radioactive waste measurement applications. It is due to be installed at Trawsfynydd NPS later this year. This paper describes the Waste Tray Assay System (WTAS) that has been developed for the measurement of Magnox FED waste. The WTAS has been tested with a range of radioactive sources and its operation has been simulated with benchmarked MCNP Monte Carlo calculations. The measurement software has been validated as has the operation of the system for a range of strong radioactive sources. A system based on the design is due for installation and operation in 2012. The system has application to the measurement of Magnox Fuel Element Debris (FED) waste at other Magnox reactor sites. The major design objective of the WTAS that has been achieved is the ability of the assay system to determine the content of Cs-137, and in turn to enable the fissile burden to be assessed using a radionuclide fingerprint, in the presence of higher and highly variable quantities of Co-60, typically from nimonic springs. The approach can be used in other Magnox FED waste configurations where the detector is located above the FED waste sorting tray and where the collimation is fixed below the detector and at a distance above the tray. In this case, which has also been investigated, there are different shielding problems and mechanical support issues. The extensive use of MCNP Monte Carlo modelling to simulate the geometry of the sorting cell and the distribution of radioactive sources has helped to ensure that all of the detector shielding requirements are addressed and suitable Cs-137 and Co-60 discrimination can be achieved. The WTAS in its present form or in other configurations has relevance to the measurement of other active ILW and highly active RH waste. Examples include high activity RH LLW and RH TRU (Transuranic) waste as defined in the United States arising from both commercial nuclear and Department of Energy (DOE) operations. The analysis is able to analyse a range of radionuclides beyong those expected in the Magnox FED cases. (authors)« less

Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

NASA Technical Reports Server (NTRS)

Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

1974-01-01

The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

LCRE and SNAP 50-DR-1 programs. Engineering progress report, January 1, 1963--March 31, 1963

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, primary coolant circuit, aaxiliary systems, fuel elements, instrumentation, materials development, and fabrication; and SNAP-50DR-1 specifications, fuel elements, pumps, steam generator, and materials development. (DCC)

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT FRAME. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-4. INEL INDEX CODE NUMBER: 075 0701 60 851 151978. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

High temperature ceramic composition for hydrogen retention

DOEpatents

Webb, R.W.

1974-01-01

A ceramic coating for H retention in fuel elements is described. The coating has relatively low thermal neutron cross section, is not readily reduced by H at 1500 deg F, is adherent to the fuel element base metal, and is stable at reactor operating temperatures. (JRD)

JACKETED URANIUM SLUG

DOEpatents

Ohlinger, L.A.; Cooper, C.M.

1958-10-01

Fuel elements for nuclear reactors are described. Eacb fuel element is comprised of a solid cylindrical slug containing fissionable material enclosed within a fluid tight jacket of neutron permeable material such as aluminum. The jacket is provided with a flexible end cap and with a sealing member having a substantially fluid-tight fit within the jacket in tight abutment with the end cap and the end of the slug. A fluid passage is provided between the end of the slug and the cap whereby leakage fiuid is principally directed to the end of the slug. In this manner, any reaction between the fissionable material and fiuid which may take place occurs more rapidly at the end of the slug than along the sides between the slug and the jacket, thereby causing longitudinal expansion of the fuel element prior to radial expansion. The longitudinal expansion can be readily detected and the fuel element removed from the coolant tube before radial expansion causes it to become jammed in the tube.

Motion analysis study on sensitivity of finite element model of the cervical spine to geometry.

PubMed

Zafarparandeh, Iman; Erbulut, Deniz U; Ozer, Ali F

2016-07-01

Numerous finite element models of the cervical spine have been proposed, with exact geometry or with symmetric approximation in the geometry. However, few researches have investigated the sensitivity of predicted motion responses to the geometry of the cervical spine. The goal of this study was to evaluate the effect of symmetric assumption on the predicted motion by finite element model of the cervical spine. We developed two finite element models of the cervical spine C2-C7. One model was based on the exact geometry of the cervical spine (asymmetric model), whereas the other was symmetric (symmetric model) about the mid-sagittal plane. The predicted range of motion of both models-main and coupled motions-was compared with published experimental data for all motion planes under a full range of loads. The maximum differences between the asymmetric model and symmetric model predictions for the principal motion were 31%, 78%, and 126% for flexion-extension, right-left lateral bending, and right-left axial rotation, respectively. For flexion-extension and lateral bending, the minimum difference was 0%, whereas it was 2% for axial rotation. The maximum coupled motions predicted by the symmetric model were 1.5° axial rotation and 3.6° lateral bending, under applied lateral bending and axial rotation, respectively. Those coupled motions predicted by the asymmetric model were 1.6° axial rotation and 4° lateral bending, under applied lateral bending and axial rotation, respectively. In general, the predicted motion response of the cervical spine by the symmetric model was in the acceptable range and nonlinearity of the moment-rotation curve for the cervical spine was properly predicted. © IMechE 2016.

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

NASA Astrophysics Data System (ADS)

Williams, W. J.; Robinson, A. B.; Rabin, B. H.

2017-12-01

This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

Chemical Dissolution of Simulant FCA Cladding and Plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Pierce, R.; O'Rourke, P.

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO 3-KF) flowsheets ofmore » H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.« less

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

Fuel cell generator energy dissipator

DOEpatents

Veyo, Stephen Emery; Dederer, Jeffrey Todd; Gordon, John Thomas; Shockling, Larry Anthony

2000-01-01

An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

SHAPED FISSIONABLE METAL BODIES

DOEpatents

Wigner, E.P.; Williamson, R.R.; Young, G.J.

1958-10-14

A technique is presented for grooving the surface of fissionable fuel elements so that expansion can take place without damage to the interior structure of the fuel element. The fissionable body tends to develop internal stressing when it is heated internally by the operation of the nuclear reactor and at the same time is subjected to surface cooling by the circulating coolant. By producing a grooved or waffle-like surface texture, the annular lines of tension stress are disrupted at equally spaced intervals by the grooves, thereby relieving the tension stresses in the outer portions of the body while also facilitating the removal of accumulated heat from the interior portion of the fuel element.

FUEL ELEMENT INTERLOCKING ARRANGEMENT

DOEpatents

Fortescue, P.; Nicoll, D.

1963-01-01

This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.; Kopelman, B.; Hausner, H.H.

1963-07-01

A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Hauth, J.J.; Anicetti, R.J.

1962-12-01

A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

FOIL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Noland, R.A.; Walker, D.E.; Spinrad, B.I.

1963-07-16

A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

The EPA National Fuels Surveillance Network. I. Trace constituents in gasoline and commercial gasoline fuel additives.

PubMed Central

Jungers, R H; Lee, R E; von Lehmden, D J

1975-01-01

A National Fuels Surveillance Network has been established to collect gasoline and other fuels through the 10 regional offices of the Environmental Protection Agency. Physical, chemical, and trace element analytical determinations are made on the collected fuel samples to detect components which may present an air pollution hazard or poison exhaust catalytic control devices. A summary of trace elemental constituents in over 50 gasoline samples and 18 commercially marketed consumer purchased gasoline additives is presented. Quantities of Mn, Ni, Cr, Zn, Cu, Fe, Sb, B, Mg, Pb, and S were found in most regular and premium gasoline. Environmental implications of trace constituents in gasoline are discussed. PMID:1157783

Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

NASA Astrophysics Data System (ADS)

Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

2017-07-01

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

Numerical Investigation of the Effect of Radial Lip Seal Geometry on Sealing Performance

NASA Astrophysics Data System (ADS)

Tok, G.; Parlar, Z.; Temiz, V.

2018-01-01

Sealing elements are often needed in industry and especially in machine design. With the change and development of machine technology from day to day, sealing elements show continuous development and change in parallel with these developments. Many factors influence the performance of the sealing elements such as shaft surface roughness, radial force, lip geometry etc. In addition, the radial lip seals must have a certain pre-load and interference in order to provide a good sealing. This also affects the friction torque. Researchers are developing new seal designs to reduce friction losses in mechanical systems. In the presented study, the effect of the lip seal geometry on sealing performance will be examined numerically. The numerical model created for this purpose will be verified with experimental data firstly. In the numerical model, shaft and seal will be modeled as hyper-elastic in 2D and 3D. NBR (Nitrile Butadiene Rubber) as seal material will be analyzed for the rotating shaft state at constant speed by applying a uniform radial force.

SODIUM DEUTERIUM REACTOR

DOEpatents

Oppenheimer, E.D.; Weisberg, R.A.

1963-02-26

This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

2011-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

Axially staggered seed-blanket reactor fuel module construction

DOEpatents

Cowell, Gary K.; DiGuiseppe, Carl P.

1985-01-01

A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-07-11

Nuclear reactor fuel elements of the type in which the flssionsble material is in ceramic form, such as uranium dioxide, are described. The fuel element is comprised of elongated inner and outer concentric spaced tubular members providing an annular space therebetween for receiving the fissionable material, the annular space being closed at both ends and the inner tube being open at both ends. The fuel is in the form of compressed pellets of ceramic fissionsble material having the configuration of split bushings formed with wedge surfaces and arranged in seriated inner and outer concentric groups which are urged against the respective tubes in response to relative axial movement of the pellets in the direction toward each other. The pairs of pellets are axially urged together by a resilient means also enclosed within the annulus. This arrangement-permits relative axial displacement of the pellets during use dial stresses on the inner and outer tube members and yet maintains the fuel pellets in good thermal conductive relationship therewith.

Autothermal reforming catalyst having perovskite structure

DOEpatents

Krumpel, Michael [Naperville, IL; Liu, Di-Jia [Naperville, IL

2009-03-24

The invention addressed two critical issues in fuel processing for fuel cell application, i.e. catalyst cost and operating stability. The existing state-of-the-art fuel reforming catalyst uses Rh and platinum supported over refractory oxide which add significant cost to the fuel cell system. Supported metals agglomerate under elevated temperature during reforming and decrease the catalyst activity. The catalyst is a perovskite oxide or a Ruddlesden-Popper type oxide containing rare-earth elements, catalytically active firs row transition metal elements, and stabilizing elements, such that the catalyst is a single phase in high temperature oxidizing conditions and maintains a primarily perovskite or Ruddlesden-Popper structure under high temperature reducing conditions. The catalyst can also contain alkaline earth dopants, which enhance the catalytic activity of the catalyst, but do not compromise the stability of the perovskite structure.

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

On the formation of string cavitation inside fuel injectors

NASA Astrophysics Data System (ADS)

Reid, B. A.; Gavaises, M.; Mitroglou, N.; Hargrave, G. K.; Garner, C. P.; Long, E. J.; McDavid, R. M.

2014-01-01

The formation of vortex or `string' cavitation has been visualised in the flow upstream of the injection hole inlet of an automotive-sized optical diesel fuel injector nozzle operating at pressures up to 2,000 bar. Three different nozzle geometries and three-dimensional flow simulations have been employed to describe how, for two adjacent nozzle holes, their relative positions influenced the formation and hole-to-hole interaction of the observed string cavitation vortices. Each hole was shown to contain two counter-rotating vortices: the first extending upstream on axis with the nozzle hole into the nozzle sac volume and the second forming a single `bridging' string linked to the adjacent hole. Steady-state and transient fuel injection conditions were shown to produce significantly different nozzle-flow characteristics with regard to the formation and interaction of these vortices in the geometries tested, with good agreement between the experimental and simulation results being achieved. The study further confirms that the visualised vortices do not cavitate themselves but act as carriers of gas-phase components within the injector flow.

Experimental investigation of piloted flameholders

NASA Technical Reports Server (NTRS)

Guo, C. F.; Zhang, Y. H.; Xie, Q. M.

1986-01-01

Four configurations of piloted flameholders were tested. The range of flame stabilization, flame propagation, pressure oscillation during ignition, and pressure drop of the configurations were determined. Some tests showed a very strong effect of inlet flow velocity profile and flameholder geometry on flame stabilization. These tests led to the following conclusions. (1) The use of a piloted flameholder in the turbofan augmentor may minimize the peak pressure rise during ignition. At the present experimental conditions, delta P/P asterisk over 2 is less than 10 percent; therefore, the use of a piloted flameholder is a good method to realize soft ignition. (2) The geometry of the piloted flameholder and the amount of fuel injected into the flameholder have a strong effect on the pressure oscillation during ignition of the fuel-air mixture in the secondary zone. (3) Compared with the V-gutter flameholder with holes in its wall, the V-gutter flameholder without holes not only has advantages such as simple structure and good rigidity but offers a wide combustion stability limit and a high capability of igniting the fuel-air mixture of the secondary zone.

BUMPERII - DESIGN ANALYSIS CODE FOR OPTIMIZING SPACECRAFT SHIELDING AND WALL CONFIGURATION FOR ORBITAL DEBRIS AND METEOROID IMPACTS

NASA Technical Reports Server (NTRS)

Hill, S. A.

1994-01-01

BUMPERII is a modular program package employing a numerical solution technique to calculate a spacecraft's probability of no penetration (PNP) from man-made orbital debris or meteoroid impacts. The solution equation used to calculate the PNP is based on the Poisson distribution model for similar analysis of smaller craft, but reflects the more rigorous mathematical modeling of spacecraft geometry, orientation, and impact characteristics necessary for treatment of larger structures such as space station components. The technique considers the spacecraft surface in terms of a series of flat plate elements. It divides the threat environment into a number of finite cases, then evaluates each element of each threat. The code allows for impact shielding (shadowing) of one element by another in various configurations over the spacecraft exterior, and also allows for the effects of changing spacecraft flight orientation and attitude. Four main modules comprise the overall BUMPERII package: GEOMETRY, RESPONSE, SHIELD, and CONTOUR. The GEOMETRY module accepts user-generated finite element model (FEM) representations of the spacecraft geometry and creates geometry databases for both meteoroid and debris analysis. The GEOMETRY module expects input to be in either SUPERTAB Universal File Format or PATRAN Neutral File Format. The RESPONSE module creates wall penetration response databases, one for meteoroid analysis and one for debris analysis, for up to 100 unique wall configurations. This module also creates a file containing critical diameter as a function of impact velocity and impact angle for each wall configuration. The SHIELD module calculates the PNP for the modeled structure given exposure time, operating altitude, element ID ranges, and the data from the RESPONSE and GEOMETRY databases. The results appear in a summary file. SHIELD will also determine the effective area of the components and the overall model, and it can produce a data file containing the probability of penetration values per surface area for each element in the model. The SHIELD module writes this data file in either SUPERTAB Universal File Format or PATRAN Neutral File Format so threat contour plots can be generated as a post-processing feature of the FEM programs SUPERTAB and PATRAN. The CONTOUR module combines the functions of the RESPONSE module and most of the SHIELD module functions allowing determination of ranges of PNP's by looping over ranges of shield and/or wall thicknesses. A data file containing the PNP's for the corresponding shield and vessel wall thickness is produced. Users may perform sensitivity studies of two kinds. The effects of simple variations in orbital time, surface area, and flux may be analyzed by making changes to the terms in the equation representing the average number of penetrating particles per unit time in the PNP solution equation. The package analyzes other changes, including model environment, surface area, and configuration, by re-running the solution sequence with new GEOMETRY and RESPONSE data. BUMPERII can be run with no interactive output to the screen during execution. This can be particularly useful during batch runs. BUMPERII is written in FORTRAN 77 for DEC VAX series computers running under VMS, and was written for use with the finite-element model code SUPERTAB or PATRAN as both a pre-processor and a post-processor. Use of an alternate FEM code will require either development of a translator to change data format or modification of the GEOMETRY subroutine in BUMPERII. This program is available in DEC VAX BACKUP format on a 9-track 1600 BPI magnetic tape (standard distribution media) or on TK50 tape cartridge. The original BUMPER code was developed in 1988 with the BUMPERII revisions following in 1991 and 1992. SUPERTAB is a former name for I-DEAS. I-DEAS Finite Element Modeling is a trademark of Structural Dynamics Research Corporation. DEC, VAX, VMS and TK50 are trademarks of Digital Equipment Corporation.

Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

Geometric k-nearest neighbor estimation of entropy and mutual information

NASA Astrophysics Data System (ADS)

Lord, Warren M.; Sun, Jie; Bollt, Erik M.

2018-03-01

Nonparametric estimation of mutual information is used in a wide range of scientific problems to quantify dependence between variables. The k-nearest neighbor (knn) methods are consistent, and therefore expected to work well for a large sample size. These methods use geometrically regular local volume elements. This practice allows maximum localization of the volume elements, but can also induce a bias due to a poor description of the local geometry of the underlying probability measure. We introduce a new class of knn estimators that we call geometric knn estimators (g-knn), which use more complex local volume elements to better model the local geometry of the probability measures. As an example of this class of estimators, we develop a g-knn estimator of entropy and mutual information based on elliptical volume elements, capturing the local stretching and compression common to a wide range of dynamical system attractors. A series of numerical examples in which the thickness of the underlying distribution and the sample sizes are varied suggest that local geometry is a source of problems for knn methods such as the Kraskov-Stögbauer-Grassberger estimator when local geometric effects cannot be removed by global preprocessing of the data. The g-knn method performs well despite the manipulation of the local geometry. In addition, the examples suggest that the g-knn estimators can be of particular relevance to applications in which the system is large, but the data size is limited.

Single shaft automotive gas turbine engine characterization test

NASA Technical Reports Server (NTRS)

Johnson, R. A.

1979-01-01

An automotive gas turbine incorporating a single stage centrifugal compressor and a single stage radial inflow turbine is described. Among the engine's features is the use of wide range variable geometry at the inlet guide vanes, the compressor diffuser vanes, and the turbine inlet vanes to achieve improved part load fuel economy. The engine was tested to determine its performance in both the variable geometry and equivalent fixed geometry modes. Testing was conducted without the originally designed recuperator. Test results were compared with the predicted performance of the nonrecuperative engine based on existing component rig test maps. Agreement between test results and the computer model was achieved.

Structural weights analysis of advanced aerospace vehicles using finite element analysis

NASA Technical Reports Server (NTRS)

Bush, Lance B.; Lentz, Christopher A.; Rehder, John J.; Naftel, J. Chris; Cerro, Jeffrey A.

1989-01-01

A conceptual/preliminary level structural design system has been developed for structural integrity analysis and weight estimation of advanced space transportation vehicles. The system includes a three-dimensional interactive geometry modeler, a finite element pre- and post-processor, a finite element analyzer, and a structural sizing program. Inputs to the system include the geometry, surface temperature, material constants, construction methods, and aerodynamic and inertial loads. The results are a sized vehicle structure capable of withstanding the static loads incurred during assembly, transportation, operations, and missions, and a corresponding structural weight. An analysis of the Space Shuttle external tank is included in this paper as a validation and benchmark case of the system.

A Methodology for the Derivation of Unloaded Abdominal Aortic Aneurysm Geometry With Experimental Validation

PubMed Central

Chandra, Santanu; Gnanaruban, Vimalatharmaiyah; Riveros, Fabian; Rodriguez, Jose F.; Finol, Ender A.

2016-01-01

In this work, we present a novel method for the derivation of the unloaded geometry of an abdominal aortic aneurysm (AAA) from a pressurized geometry in turn obtained by 3D reconstruction of computed tomography (CT) images. The approach was experimentally validated with an aneurysm phantom loaded with gauge pressures of 80, 120, and 140 mm Hg. The unloaded phantom geometries estimated from these pressurized states were compared to the actual unloaded phantom geometry, resulting in mean nodal surface distances of up to 3.9% of the maximum aneurysm diameter. An in-silico verification was also performed using a patient-specific AAA mesh, resulting in maximum nodal surface distances of 8 μm after running the algorithm for eight iterations. The methodology was then applied to 12 patient-specific AAA for which their corresponding unloaded geometries were generated in 5–8 iterations. The wall mechanics resulting from finite element analysis of the pressurized (CT image-based) and unloaded geometries were compared to quantify the relative importance of using an unloaded geometry for AAA biomechanics. The pressurized AAA models underestimate peak wall stress (quantified by the first principal stress component) on average by 15% compared to the unloaded AAA models. The validation and application of the method, readily compatible with any finite element solver, underscores the importance of generating the unloaded AAA volume mesh prior to using wall stress as a biomechanical marker for rupture risk assessment. PMID:27538124

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

Differential die-away instrument: Report on comparison of fuel assembly experiments and simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Henzl, Vladimir; Swinhoe, Martyn Thomas

2015-01-14

Experimental results of the assay of mock-up (fresh) fuel with the differential die-away (DDA) instrument were compared to the Monte Carlo N-Particle eXtended (MCNPX) simulation results. Most principal experimental observables, the die-away time and the in tegral of the DDA signal in several time domains, have been found in good agreement with the MCNPX simulation results. The remaining discrepancies between the simulation and experimental results are likely due to small differences between the actual experimental setup and the simulated geometry, including uncertainty in the DT neutron generator yield. Within this report we also present a sensitivity study of the DDAmore » instrument which is a complex and sensitive system and demonstrate to what degree it can be impacted by geometry, material composition, and electronics performance.« less

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

Production test IP-544-A, irradiation of 1.6% enriched thick walled single tube elements in KER-1 and 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kratzer, W.K.; Wise, M.J.

1962-12-12

The objective of this production test is to authorize the irradiation of coextruded Zr-2 jacketed thick walled 1.6% enriched tubular elements in KER loops 1 and 2 to evaluate the swelling behavior of fuel elements at high uranium temperatures Coextruded Zr-2 jacketed 1.6% enriched tubular fuel elements 1.79 inch OD, 0.97 inch ID, and 12 inches long will be irradiated KER loops 1 and 2 to exposures no greater than 2500 MWD/T.

Fuel conservation possibilities for terminal area compatible aircraft

NASA Technical Reports Server (NTRS)

1975-01-01

Design features and operational procedures are identified, which would reduce fuel consumption of future transport aircraft. The fuel-saving potential can be realized during the last decade of this century only if the necessary research and technology programs are implemented in the areas of composite primary structure, airfoil/wing design, and stability augmentation systems. The necessary individual R and T programs are defined. The sensitivity to fuel usage of several design parameters (wing geometry, cruise speed, propulsion) is investigated, and the results applied to a candidate 18, 140-kg (40,000-lb) payload, 5556-km (3000-nmi) transport design. Technical and economic comparisons are made with current commercial aircraft and other advanced designs.

Validation of MCNP: SPERT-D and BORAX-V fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Validation of MCNP: SPERT-D and BORAX-V fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Redwing: A MOOSE application for coupling MPACT and BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frederick N. Gleicher; Michael Rose; Tom Downar

Fuel performance and whole core neutron transport programs are often used to analyze fuel behavior as it is depleted in a reactor. For fuel performance programs, internal models provide the local intra-pin power density, fast neutron flux, burnup, and fission rate density, which are needed for a fuel performance analysis. The fuel performance internal models have a number of limitations. These include effects on the intra-pin power distribution by nearby assembly elements, such as water channels and control rods, and the further limitation of applicability to a specified fuel type such as low enriched UO2. In addition, whole core neutronmore » transport codes need an accurate intra-pin temperature distribution in order to calculate neutron cross sections. Fuel performance simulations are able to model the intra-pin fuel displacement as the fuel expands and densifies. These displacements must be accurately modeled in order to capture the eventual mechanical contact of the fuel and the clad; the correct radial gap width is needed for an accurate calculation of the temperature distribution of the fuel rod. Redwing is a MOOSE-based application that enables coupling between MPACT and BISON for transport and fuel performance coupling. MPACT is a 3D neutron transport and reactor core simulator based on the method of characteristics (MOC). The development of MPACT began at the University of Michigan (UM) and now is under the joint development of ORNL and UM as part of the DOE CASL Simulation Hub. MPACT is able to model the effects of local assembly elements and is able calculate intra-pin quantities such as the local power density on a volumetric mesh for any fuel type. BISON is a fuel performance application of Multi-physics Object Oriented Simulation Environment (MOOSE), which is under development at Idaho National Laboratory. BISON is able to solve the nonlinearly coupled mechanical deformation and heat transfer finite element equations that model a fuel element as it is depleted in a nuclear reactor. Redwing couples BISON and MPACT in a single application. Redwing maps and transfers the individual intra-pin quantities such as fission rate density, power density, and fast neutron flux from the MPACT volumetric mesh to the individual BISON finite element meshes. For a two-way coupling Redwing maps and transfers the individual pin temperature field and axially dependent coolant densities from the BISON mesh to the MPACT volumetric mesh. Details of the mapping are given. Redwing advances the simulation with the MPACT solution for each depletion time step and then advances the multiple BISON simulations for fuel performance calculations. Sub-cycle advancement can be applied to the individual BISON simulations and allows multiple time steps to be applied to the fuel performance simulations. Currently, only loose coupling where data from a previous time step is applied to the current time step is performed.« less

Rolling-Element Bearings

NASA Technical Reports Server (NTRS)

Hamrock, B. J.; Anderson, W. J.

1983-01-01

Rolling element bearings are a precision, yet simple, machine element of great utility. A brief history of rolling element bearings is reviewed and the type of rolling element bearings, their geometry and kinematics, as well as the materials they are made from and the manufacturing processes they involve are described. Unloaded and unlubricated rolling element bearings, loaded but unlubricated rolling element bearings and loaded and lubricated rolling element bearings are considered. The recognition and understanding of elastohydrodynamic lubrication covered, represents one of the major development in rolling element bearings.

Stem geometry changes initial femoral fixation stability of a revised press-fit hip prosthesis: A finite element study.

PubMed

Russell, Robert D; Huo, Michael H; Rodrigues, Danieli C; Kosmopoulos, Victor

2016-11-14

Stable femoral fixation during uncemented total hip arthroplasty is critical to allow for subsequent osseointegration of the prosthesis. Varying stem designs provide surgeons with multiple options to gain femoral fixation. The purpose of this study was to compare the initial fixation stability of cylindrical and tapered stem implants using two different underreaming techniques (press-fit conditions) for revision total hip arthroplasty (THA). A finite element femur model was created from three-dimensional computed tomography images simulating a trabecular bone defect commonly observed in revision THA. Two 18-mm generic femoral hip implants were modeled using the same geometry, differing only in that one had a cylindrical stem and the other had a 2 degree tapered stem. Surgery was simulated using a 0.05-mm and 0.01-mm press-fit and tested with a physiologically relevant loading protocol. Mean contact pressure was influenced more by the surgical technique than by the stem geometry. The 0.05-mm press-fit condition resulted in the highest contact pressures for both the cylindrical (27.35 MPa) and tapered (20.99 MPa) stems. Changing the press-fit to 0.01-mm greatly decreased the contact pressure by 79.8% and 78.5% for the cylindrical (5.53 MPa) and tapered (4.52 MPa) models, respectively. The cylindrical stem geometry consistently showed less relative micromotion at all the cross-sections sampled as compared to the tapered stem regardless of press-fit condition. This finite element analysis study demonstrates that tapered stem results in lower average contact pressure and greater micromotion at the implant-bone interface than a cylindrical stem geometry. More studies are needed to establish how these different stem geometries perform in such non-ideal conditions encountered in revision THA cases where less bone stock is available.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, K.C.

1988-01-21

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, Kenny C.

1989-01-01

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

Investigation of High Pressure, Multi-Hole Diesel Fuel Injection Using High Speed Imaging

NASA Astrophysics Data System (ADS)

Morris, Steven; Eagle, Ethan; Wooldridge, Margaret

2012-10-01

Research to experimentally capture and understand transient fuel spray behavior of modern fuel injection systems remains underdeveloped. To this end, a high-pressure diesel common-rail fuel injector was instrumented in a spherical, constant volume combustion chamber to image the early time history of injection of diesel fuel. The research-geometry fuel injector has four holes aligned on a radial plane of the nozzle with hole sizes of 90, 110, 130 and 150 μm in diameter. Fuel was injected into a non-reacting environment with ambient densities of 17.4, 24.0, and 31.8 kg/m3 at fuel rail pressures of 1000, 1500, and 2000 bar. High speed images of fuel injection were taken using backlighting at 100,000 frames per second (100 kfps) and an image processing algorithm. The experimental results are compared with a one-dimensional fuel-spray model that was historically developed and applied to fuel sprays from single-hole fuel injectors. Fuel spray penetration distance was evaluated as a function of time for the different injector hole diameters, fuel injection pressures and ambient densities. The results show the differences in model predictions and experimental data at early times in the spray development.

NEUTRONIC REACTOR CONSTRUCTION

DOEpatents

Vernon, H.C.; Goett, J.J.

1958-09-01

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

TWISTED RIBBON FUEL ELEMENT

DOEpatents

Breden, C.R.; Schultz, A.B.

1961-06-01

A reactor core formed of bundles of parallel fuel elements in the form of ribbons is patented. The fuel ribbons are twisted about their axes so as to have contact with one another at regions spaced lengthwise of the ribbons and to be out of contact with one another at locations between these spaced regions. The contact between the ribbons is sufficient to allow them to be held together in a stable bundle in a containing tube without intermediate support, while permitting enough space between the ribbon for coolant flowing.

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

Alternate fusion fuels workshop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-06-01

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

Nitsche Extended Finite Element Methods for Earthquake Simulation

NASA Astrophysics Data System (ADS)

Coon, Ethan T.

Modeling earthquakes and geologically short-time-scale events on fault networks is a difficult problem with important implications for human safety and design. These problems demonstrate a. rich physical behavior, in which distributed loading localizes both spatially and temporally into earthquakes on fault systems. This localization is governed by two aspects: friction and fault geometry. Computationally, these problems provide a stern challenge for modelers --- static and dynamic equations must be solved on domains with discontinuities on complex fault systems, and frictional boundary conditions must be applied on these discontinuities. The most difficult aspect of modeling physics on complicated domains is the mesh. Most numerical methods involve meshing the geometry; nodes are placed on the discontinuities, and edges are chosen to coincide with faults. The resulting mesh is highly unstructured, making the derivation of finite difference discretizations difficult. Therefore, most models use the finite element method. Standard finite element methods place requirements on the mesh for the sake of stability, accuracy, and efficiency. The formation of a mesh which both conforms to fault geometry and satisfies these requirements is an open problem, especially for three dimensional, physically realistic fault. geometries. In addition, if the fault system evolves over the course of a dynamic simulation (i.e. in the case of growing cracks or breaking new faults), the geometry must he re-meshed at each time step. This can be expensive computationally. The fault-conforming approach is undesirable when complicated meshes are required, and impossible to implement when the geometry is evolving. Therefore, meshless and hybrid finite element methods that handle discontinuities without placing them on element boundaries are a desirable and natural way to discretize these problems. Several such methods are being actively developed for use in engineering mechanics involving crack propagation and material failure. While some theory and application of these methods exist, implementations for the simulation of networks of many cracks have not yet been considered. For my thesis, I implement and extend one such method, the eXtended Finite Element Method (XFEM), for use in static and dynamic models of fault networks. Once this machinery is developed, it is applied to open questions regarding the behavior of networks of faults, including questions of distributed deformation in fault systems and ensembles of magnitude, location, and frequency in repeat ruptures. The theory of XFEM is augmented to allow for solution of problems with alternating regimes of static solves for elastic stress conditions and short, dynamic earthquakes on networks of faults. This is accomplished using Nitsche's approach for implementing boundary conditions. Finally, an optimization problem is developed to determine tractions along the fault, enabling the calculation of frictional constraints and the rupture front. This method is verified via a series of static, quasistatic, and dynamic problems. Armed with this technique, we look at several problems regarding geometry within the earthquake cycle in which geometry is crucial. We first look at quasistatic simulations on a community fault model of Southern California, and model slip distribution across that system. We find the distribution of deformation across faults compares reasonably well with slip rates across the region, as constrained by geologic data. We find geometry can provide constraints for friction, and consider the minimization of shear strain across the zone as a function of friction and plate loading direction, and infer bounds on fault strength in the region. Then we consider the repeated rupture problem, modeling the full earthquake cycle over the course of many events on several fault geometries. In this work, we look at distributions of events, studying the effect of geometry on statistical metrics of event ensembles. Finally, this thesis is a proof of concept for the XFEM on earthquake cycle models on fault systems. We identify strengths and weaknesses of the method, and identify places for future improvement. We discuss the feasibility of the method's use in three dimensions, and find the method to be a strong candidate for future crustal deformation simulations.

Optimizing detector geometry for trace element mapping by X-ray fluorescence

DOE PAGES

Sun, Yue; Gleber, Sophie -Charlotte; Jacobsen, Chris; ...

2015-01-01

We report that trace metals play critical roles in a variety of systems, ranging from cells to photovoltaics. X-Ray Fluorescence (XRF) microscopy using X-ray excitation provides one of the highest sensitivities available for imaging the distribution of trace metals at sub-100 nm resolution. With the growing availability and increasing performance of synchrotron light source based instruments and X-ray nanofocusing optics, and with improvements in energy-dispersive XRF detectors, what are the factors that limit trace element detectability? To address this question, we describe an analytical model for the total signal incident on XRF detectors with various geometries, including the spectral responsemore » of energy dispersive detectors. This model agrees well with experimentally recorded X-ray fluorescence spectra, and involves much shorter calculation times than with Monte Carlo simulations. With such a model, one can estimate the signal when a trace element is illuminated with an X-ray beam, and when just the surrounding non-fluorescent material is illuminated. From this signal difference, a contrast parameter can be calculated and this can in turn be used to calculate the signal-to-noise ratio (S/N) for detecting a certain elemental concentration. We apply this model to the detection of trace amounts of zinc in biological materials, and to the detection of small quantities of arsenic in semiconductors. In conclusion, we conclude that increased detector collection solid angle is (nearly) always advantageous even when considering the scattered signal. However, given the choice between a smaller detector at 90° to the beam versus a larger detector at 180° (in a backscatter-like geometry), the 90° detector is better for trace element detection in thick samples, while the larger detector in 180° geometry is better suited to trace element detection in thin samples.« less

Optimizing detector geometry for trace element mapping by X-ray fluorescence

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Yue; Gleber, Sophie -Charlotte; Jacobsen, Chris

We report that trace metals play critical roles in a variety of systems, ranging from cells to photovoltaics. X-Ray Fluorescence (XRF) microscopy using X-ray excitation provides one of the highest sensitivities available for imaging the distribution of trace metals at sub-100 nm resolution. With the growing availability and increasing performance of synchrotron light source based instruments and X-ray nanofocusing optics, and with improvements in energy-dispersive XRF detectors, what are the factors that limit trace element detectability? To address this question, we describe an analytical model for the total signal incident on XRF detectors with various geometries, including the spectral responsemore » of energy dispersive detectors. This model agrees well with experimentally recorded X-ray fluorescence spectra, and involves much shorter calculation times than with Monte Carlo simulations. With such a model, one can estimate the signal when a trace element is illuminated with an X-ray beam, and when just the surrounding non-fluorescent material is illuminated. From this signal difference, a contrast parameter can be calculated and this can in turn be used to calculate the signal-to-noise ratio (S/N) for detecting a certain elemental concentration. We apply this model to the detection of trace amounts of zinc in biological materials, and to the detection of small quantities of arsenic in semiconductors. In conclusion, we conclude that increased detector collection solid angle is (nearly) always advantageous even when considering the scattered signal. However, given the choice between a smaller detector at 90° to the beam versus a larger detector at 180° (in a backscatter-like geometry), the 90° detector is better for trace element detection in thick samples, while the larger detector in 180° geometry is better suited to trace element detection in thin samples.« less

Chemical characterization of biomass fuel smoke particles of rural kitchens of South Asia

NASA Astrophysics Data System (ADS)

Deka, Pratibha; Hoque, Raza Rafiqul

2015-05-01

Biomass fuel smoke particles (BFSPs) of rural kitchens collected during dry and wet seasons were characterized for elements, anions and carbon. The BFSPs of kitchens using varied biomass fuel types viz. cow dung stick, mixed biomass, cow-dung stick-mixed biomass and sugarcane bagasse were chosen for the study. The BFSPs from cow dung fuel stick showed higher levels of elements, anions and particulate carbon than other BFSPs. Calcium, K, Fe and Mg were the major elements found in all BFSPs, which did not vary much between the seasons. Sulphate was found to be the dominant anion present in all BFSPs followed by Clˉ and PO43-. Seasonal variation was pronounced in the case of abundance of anions and particulate carbon. The ratio OC/EC, often used as source signature of biomass burning, was found to be within 1.89-7.41 and 1.72-6.19 during dry and wet seasons respectively.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

ON CRITICAL MASS ANALYSIS OF JRR-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1961-01-01

The critica mass of the JRR-2 was found to be 15 fuel elements, instead of 8 as expected, when the reactor reached criticaity. The critica mass was analyzed by AMF and JAERI a few years ago, but afterwards some modifications have been made of the stucture for the reinforcement, for example, during the construction. The critical mass is recalculated perfectly and the difference bctween 15 and S fuel elements is discussed. The deviation of the critical mass is mainly caused by the effects of control rods, fuel elcments, grid-plate, etc., in the reflector; only heavy water or light water wasmore » conaidered as the reflector in the previous calculation. A simple method is used to calculate the critical mass. The effective multiplication factor for the core with 15 fuel elements is obtained about 2% higher than the experimental value. This difference is also discussed in detail. (auth)« less

Modeling and Simulation of a Nuclear Fuel Element Test Section

NASA Technical Reports Server (NTRS)

Moran, Robert P.; Emrich, William

2011-01-01

"The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

A geometry package for generation of input data for a three-dimensional potential-flow program

NASA Technical Reports Server (NTRS)

Halsey, N. D.; Hess, J. L.

1978-01-01

The preparation of geometric data for input to three-dimensional potential flow programs was automated and simplified by a geometry package incorporated into the NASA Langley version of the 3-D lifting potential flow program. Input to the computer program for the geometry package consists of a very sparse set of coordinate data, often with an order of magnitude of fewer points than required for the actual potential flow calculations. Isolated components, such as wings, fuselages, etc. are paneled automatically, using one of several possible element distribution algorithms. Curves of intersection between components are calculated, using a hybrid curve-fit/surface-fit approach. Intersecting components are repaneled so that adjacent elements on either side of the intersection curves line up in a satisfactory manner for the potential-flow calculations. Many cases may be run completely (from input, through the geometry package, and through the flow calculations) without interruption. Use of the package significantly reduces the time and expense involved in making three-dimensional potential flow calculations.

Discovering Structural Regularity in 3D Geometry

PubMed Central

Pauly, Mark; Mitra, Niloy J.; Wallner, Johannes; Pottmann, Helmut; Guibas, Leonidas J.

2010-01-01

We introduce a computational framework for discovering regular or repeated geometric structures in 3D shapes. We describe and classify possible regular structures and present an effective algorithm for detecting such repeated geometric patterns in point- or mesh-based models. Our method assumes no prior knowledge of the geometry or spatial location of the individual elements that define the pattern. Structure discovery is made possible by a careful analysis of pairwise similarity transformations that reveals prominent lattice structures in a suitable model of transformation space. We introduce an optimization method for detecting such uniform grids specifically designed to deal with outliers and missing elements. This yields a robust algorithm that successfully discovers complex regular structures amidst clutter, noise, and missing geometry. The accuracy of the extracted generating transformations is further improved using a novel simultaneous registration method in the spatial domain. We demonstrate the effectiveness of our algorithm on a variety of examples and show applications to compression, model repair, and geometry synthesis. PMID:21170292

Prospects for reduced energy transports: A preliminary analysis

NASA Technical Reports Server (NTRS)

Ardema, M. D.; Harper, M.; Smith, C. L.; Waters, M. H.; Williams, L. J.

1974-01-01

The recent energy crisis and subsequent substantial increase in fuel prices have provided increased incentive to reduce the fuel consumption of civil transport aircraft. At the present time many changes in operational procedures have been introduced to decrease fuel consumption of the existing fleet. In the future, however, it may become desirable or even necessary to introduce new fuel-conservative aircraft designs. This paper reports the results of a preliminary study of new near-term fuel conservative aircraft. A parametric study was made to determine the effects of cruise Mach number and fuel cost on the optimum configuration characteristics and on economic performance. For each design, the wing geometry was optimized to give maximum return on investment at a particular fuel cost. Based on the results of the parametric study, a nominal reduced energy configuration was selected. Compared with existing transport designs, the reduced energy design has a higher aspect ratio wing with lower sweep, and cruises at a lower Mach number. It has about 30% less fuel consumption on a seat-mile basis.

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT FRAME AND SUPPORT PLATFORM, AND SAFETY MECHANISM ASSEMBLY (SPRING-LOADED HINGE). F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-1. INEL INDEX CODE NUMBER: 075 0701 60 851 151975. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... transuranic elements. Different technical processes can accomplish this separation. However, over the years Purex has become the most commonly used and accepted process. Purex involves the dissolution of... facilities have process functions similar to each other, including: irradiated fuel element chopping, fuel...

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

A New Approach to the Computer Modeling of Amorphous Nanoporous Structures of Semiconducting and Metallic Materials: A Review

PubMed Central

Romero, Cristina; Noyola, Juan C.; Santiago, Ulises; Valladares, Renela M.; Valladares, Alexander; Valladares, Ariel A.

2010-01-01

We review our approach to the generation of nanoporous materials, both semiconducting and metallic, which leads to the existence of nanopores within the bulk structure. This method, which we have named as the expanding lattice method, is a novel transferable approach which consists first of constructing crystalline supercells with a large number of atoms and a density close to the real value and then lowering the density by increasing the volume. The resulting supercells are subjected to either ab initio or parameterized—Tersoff-based—molecular dynamics processes at various temperatures, all below the corresponding bulk melting points, followed by geometry relaxations. The resulting samples are essentially amorphous and display pores along some of the “crystallographic” directions without the need of incorporating ad hoc semiconducting atomic structural elements such as graphene-like sheets and/or chain-like patterns (reconstructive simulations) or of reproducing the experimental processes (mimetic simulations). We report radial (pair) distribution functions, nanoporous structures of C and Si, and some computational predictions for their vibrational density of states. We present numerical estimates and discuss possible applications of semiconducting materials for hydrogen storage in potential fuel tanks. Nanopore structures for metallic elements like Al and Au also obtained through the expanding lattice method are reported.

Computational Investigation of Combustion Instabilities in a Laboratory-Scale LDI Gas Turbine Engine

DTIC Science & Technology

2013-06-01

combustor by the insertion of a slotted inlet and an exit nozzle , whereas the reduced geometry is acoustically open. Table 2 Summary of Cases Considered... nozzle located at the right-end surface, an outlet condition is imposed by a characteristic back pressure condition. The fuel spray is injected at the...Computational Mesh visualized around the fuel nozzle and swirler III. Decomposition Methods For Combustion Dynamics Diagnostics To understand the

NTREES Testing and Operations Status

NASA Technical Reports Server (NTRS)

Emrich, Bill

2007-01-01

Nuclear Thermal Rockets or NTR's have been suggested as a propulsion system option for vehicles traveling to the moon or Mars. These engines are capable of providing high thrust at specific impulses at least twice that of today's best chemical engines. The performance constraints on these engines are mainly the result of temperature limitations on the fuel coupled with a limited ability to withstand chemical attack by the hot hydrogen propellant. To operate at maximum efficiency, fuel forms are desired which can withstand the extremely hot, hostile environment characteristic of NTR operation for at least several hours. The simulation of such an environment would require an experimental device which could simultaneously approximate the power, flow, and temperature conditions which a nuclear fuel element (or partial element) would encounter during NTR operation. Such a simulation would allow detailed studies of the fuel behavior and hydrogen flow characteristics under reactor like conditions to be performed. Currently, the construction of such a simulator has been completed at the Marshall Space Flight Center, and will be used in the future to evaluate a wide variety of fuel element designs and the materials of which they are fabricated. This present work addresses the operational status of the Nuclear Thermal Rocket Element Environmental Simulator or NTREES and some of the design considerations which were considered prior to and during its construction.

GEH-4-42, 47; Hot pressed, I and E cooled fuel element irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neidner, R.

1959-11-02

In our continual effort to improve the present fuel elements which are irradiated in the numerous Hanford reactors, we have made what we believe to be a significant improvement in the hot pressing process for jacketing uranium fuel slugs. We are proposing a large scale evaluation testing program in the Hanford reactors but need the vital and basic information on the operating characteristics of this type slug under known and controlled operating conditions. We, therefore, have prepared two typical fuel slugs and will want them irradiated to about 1000 MWD/T exposure (this will require about four to five total cycles).

CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

DOEpatents

Duckworth, W.H.

1957-12-01

This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

Nonlinear finite element modeling of corrugated board

Treesearch

A. C. Gilchrist; J. C. Suhling; T. J. Urbanik

1999-01-01

In this research, an investigation on the mechanical behavior of corrugated board has been performed using finite element analysis. Numerical finite element models for corrugated board geometries have been created and executed. Both geometric (large deformation) and material nonlinearities were included in the models. The analyses were performed using the commercial...

Load responsive hydrodynamic bearing

DOEpatents

Kalsi, Manmohan S.; Somogyi, Dezso; Dietle, Lannie L.

2002-01-01

A load responsive hydrodynamic bearing is provided in the form of a thrust bearing or journal bearing for supporting, guiding and lubricating a relatively rotatable member to minimize wear thereof responsive to relative rotation under severe load. In the space between spaced relatively rotatable members and in the presence of a liquid or grease lubricant, one or more continuous ring shaped integral generally circular bearing bodies each define at least one dynamic surface and a plurality of support regions. Each of the support regions defines a static surface which is oriented in generally opposed relation with the dynamic surface for contact with one of the relatively rotatable members. A plurality of flexing regions are defined by the generally circular body of the bearing and are integral with and located between adjacent support regions. Each of the flexing regions has a first beam-like element being connected by an integral flexible hinge with one of the support regions and a second beam-like element having an integral flexible hinge connection with an adjacent support region. A least one local weakening geometry of the flexing region is located intermediate the first and second beam-like elements. In response to application of load from one of the relatively rotatable elements to the bearing, the beam-like elements and the local weakening geometry become flexed, causing the dynamic surface to deform and establish a hydrodynamic geometry for wedging lubricant into the dynamic interface.

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Renfro, David G; Chandler, David; Cook, David Howard

2014-11-01

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully convertedmore » using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

A collection of edge-based elements

NASA Technical Reports Server (NTRS)

Kempel, Leo C.; Volakis, John L.

1992-01-01

Edge-based elements have proved useful in solving electromagnetic problems since they are nondivergent. Previous authors have presented several two and three dimensional elements. Herein, we present four types of elements which are suitable for modeling several types of three dimensional geometries. Distorted brick and triangular prism elements are given in cartesian coordinates as well as the specialized cylindrical shell and pie-shaped prism elements which are suitable for problems best described in polar cylindrical coordinates.

NASA's Nuclear Thermal Propulsion Project

NASA Technical Reports Server (NTRS)

Houts, Michael; Mitchell, Sonny; Kim, Tony; Borowski, Stanley; Power, Kevin; Scott, John; Belvin, Anthony; Clement, Steven

2015-01-01

Space fission power systems can provide a power rich environment anywhere in the solar system, independent of available sunlight. Space fission propulsion offers the potential for enabling rapid, affordable access to any point in the solar system. One type of space fission propulsion is Nuclear Thermal Propulsion (NTP). NTP systems operate by using a fission reactor to heat hydrogen to very high temperature (>2500 K) and expanding the hot hydrogen through a supersonic nozzle. First generation NTP systems are designed to have an Isp of approximately 900 s. The high Isp of NTP enables rapid crew transfer to destinations such as Mars, and can also help reduce mission cost, improve logistics (fewer launches), and provide other benefits. However, for NTP systems to be utilized they must be affordable and viable to develop. NASA's Advanced Exploration Systems (AES) NTP project is a technology development project that will help assess the affordability and viability of NTP. Early work has included fabrication of representative graphite composite fuel element segments, coating of representative graphite composite fuel element segments, fabrication of representative cermet fuel element segments, and testing of fuel element segments in the Compact Fuel Element Environmental Tester (CFEET). Near-term activities will include testing approximately 16" fuel element segments in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES), and ongoing research into improving fuel microstructure and coatings. In addition to recapturing fuels technology, affordable development, qualification, and utilization strategies must be devised. Options such as using low-enriched uranium (LEU) instead of highly-enriched uranium (HEU) are being assessed, although that option requires development of a key technology before it can be applied to NTP in the thrust range of interest. Ground test facilities will be required, especially if NTP is to be used in conjunction with high value or crewed missions. There are potential options for either modifying existing facilities or constructing new ground test facilities. At least three potential options exist for reducing (or eliminating) the release of radioactivity into the environment during ground testing. These include fully containing the NTP exhaust during the ground test, scrubbing the exhaust, or utilizing an existing borehole at the Nevada National Security Site (NNSS) to filter the exhaust. Finally, the project is considering the potential for an early flight demonstration of an engine very similar to one that could be used to support human Mars or other ambitious missions. The flight demonstration could be an important step towards the eventual utilization of NTP.

Advanced diesel electronic fuel injection and turbocharging

NASA Astrophysics Data System (ADS)

Beck, N. J.; Barkhimer, R. L.; Steinmeyer, D. C.; Kelly, J. E.

1993-12-01

The program investigated advanced diesel air charging and fuel injection systems to improve specific power, fuel economy, noise, exhaust emissions, and cold startability. The techniques explored included variable fuel injection rate shaping, variable injection timing, full-authority electronic engine control, turbo-compound cooling, regenerative air circulation as a cold start aid, and variable geometry turbocharging. A Servojet electronic fuel injection system was designed and manufactured for the Cummins VTA-903 engine. A special Servojet twin turbocharger exhaust system was also installed. A series of high speed combustion flame photos was taken using the single cylinder optical engine at Michigan Technological University. Various fuel injection rate shapes and nozzle configurations were evaluated. Single-cylinder bench tests were performed to evaluate regenerative inlet air heating techniques as an aid to cold starting. An exhaust-driven axial cooling air fan was manufactured and tested on the VTA-903 engine.

An investigation of air solubility in Jet A fuel at high pressures

NASA Technical Reports Server (NTRS)

Faeth, G. M.

1981-01-01

Problems concerned with the supercritical injection concept are discussed. Supercritical injection involves dissolving air into a fuel prior to injection. A similar effect is obtained by preheating the fuel so that a portion of the fuel flashes when its pressure is reduced. Flashing improves atomization properties and the presence of air in the primary zone of a spray flame reduces the formation of pollutants. The investigation is divided into three phases: (1) measure the solubility and density properties of fuel/gas mixtures, including Jet A/air, at pressures and correlate these results using theory; (2) investigate the atomization properties of flashing liquids, including fuel/dissolved gas systems. Determine and correlate the effect of inlet properties and injector geometry on mass flow rates, Sauter mean diameter and spray angles; (3) examine the combustion properties of flashing injection in an open burner flame, considering flame shape and soot production.

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS, FUEL ELEMENT CUTTING FACILITY, AND DRY GRAPHITE STORAGE FACILITY. INL DRAWING NUMBER 200-0603-00-030-056329. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

40 CFR 80.1141 - Small refinery exemption.

Code of Federal Regulations, 2010 CFR

2010-07-01

... (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Renewable Fuel Standard § 80.1141 Small refinery exemption...)), is exempt from the renewable fuel standards of § 80.1105 and the requirements that apply to obligated... refinery application. The application must contain all of the elements required for small refinery...