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.

On the Nonlinear Behavior of a Glass-Ceramic Seal and its Application in Planar SOFC Systems

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

Nguyen, Ba Nghiep; Koeppel, Brian J.; Vetrano, John S.

2006-06-01

This paper studies the nonlinear behavior of a glass-ceramic seal used in planar solid oxide fuel cells (SOFCs). To this end, a viscoelastic damage model has been developed that can capture the nonlinear material response due to both progressive damage in the glass-ceramic material and viscous flow of the residual glass in this material. The model has been implemented in the MSC MARC finite element code, and its validation has been carried out using the experimental relaxation test data obtained for this material at 700oC, 750oC, and 800oC. Finally, it has been applied to the simulation of a SOFC stackmore » under thermal cycling conditions. The areas of potential damage have been predicted.« less

Thermal-structural analyses of Space Shuttle Main Engine (SSME) hot section components

NASA Technical Reports Server (NTRS)

Abdul-Aziz, Ali; Thompson, Robert L.

1988-01-01

Three dimensional nonlinear finite element heat transfer and structural analyses were performed for the first stage high pressure fuel turbopump (HPFTP) blade of the space shuttle main engine (SSME). Directionally solidified (DS) MAR-M 246 and single crystal (SC) PWA-1480 material properties were used for the analyses. Analytical conditions were based on a typical test stand engine cycle. Blade temperature and stress strain histories were calculated by using the MARC finite element computer code. The structural response of an SSME turbine blade was assessed and a greater understanding of blade damage mechanisms, convective cooling effects, and thermal mechanical effects was gained.

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.

Fuel Tank Technology

DTIC Science & Technology

1989-11-01

the high risk of fuel cells damaging as a consequence of the unfolding and refolding operations. - Difficulties to perform acceptance inspection tests...corners sometimes present in the structures. (See FIG. 6, 7, 8). - Additional installation costs and risk of damaging due to fuel cells anchoring...performed manually by very complex tying operations. (See. FIG. 9). - Risk of damaging of the thicker reinforced zones of the flexible fuel cells where

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.

75 FR 3876 - Mark Edward Leyse; Receipt of Petition for Rulemaking

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

... (assembly) severe fuel damage experiments. The petitioner also requests that the NRC promulgate a regulation... aware that data from multi-rod (assembly) severe fuel damage experiments indicates that the current... fuel damage experiments indicates that the current peak cladding temperature limit contained in 10 CFR...

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.

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.

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.

Ecological risk assessment to support fuels treatment project decisions

Treesearch

Jay O' Laughlin

2010-01-01

Risk is a combined statement of the probability that something of value will be damaged and some measure of the damage’s adverse effect. Wildfires burning in the uncharacteristic fuel conditions now typical throughout the Western United States can damage ecosystems and adversely affect environmental conditions. Wildfire behavior can be modified by prefire fuel...

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)

Three-Dimensional Analysis of Voids in AM60B Magnesium Tensile Bars Using Computed Tomography Imagery

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

Waters, A M

2001-05-01

In an effort to increase automobile fuel efficiency as well as decrease the output of harmful greenhouse gases, the automotive industry has recently shown increased interest in cast light metals such as magnesium alloys in an effort to increase weight savings. Currently several magnesium alloys such as AZ91 and AM60B are being used in structural applications for automobiles. However, these magnesium alloys are not as well characterized as other commonly used structural metals such as aluminum. This dissertation presents a methodology to nondestructively quantify damage accumulation due to void behavior in three dimensions in die-cast magnesium AM60B tensile bars asmore » a function of mechanical load. Computed tomography data was acquired after tensile bars were loaded up to and including failure, and analyzed to characterize void behavior as it relates to damage accumulation. Signal and image processing techniques were used along with a cluster labeling routine to nondestructively quantify damage parameters in three dimensions. Void analyses were performed including void volume distribution characterization, nearest neighbor distance calculations, shape parameters, and volumetric renderings of voids in the alloy. The processed CT data was used to generate input files for use in finite element simulations, both two- and three-dimensional. The void analyses revealed that the overwhelming source of failure in each tensile bar was a ring of porosity within each bar, possibly due to a solidification front inherent to the casting process. The measured damage parameters related to void nucleation, growth, and coalescence were shown to contribute significantly to total damage accumulation. Void volume distributions were characterized using a Weibull function, and the spatial distributions of voids were shown to be clustered. Two-dimensional finite element analyses of the tensile bars were used to fine-tune material damage models and a three-dimensional mesh of an extracted portion of one tensile bar including voids was generated from CT data and used as input to a finite element analysis.« less

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.)

Damage detection in composite panels based on mode-converted Lamb waves sensed using 3D laser scanning vibrometer

NASA Astrophysics Data System (ADS)

Pieczonka, Łukasz; Ambroziński, Łukasz; Staszewski, Wiesław J.; Barnoncel, David; Pérès, Patrick

2017-12-01

This paper introduces damage identification approach based on guided ultrasonic waves and 3D laser Doppler vibrometry. The method is based on the fact that the symmetric and antisymmetric Lamb wave modes differ in amplitude of the in-plane and out-of-plane vibrations. Moreover, the modes differ also in group velocities and normally they are well separated in time. For a given time window both modes can occur simultaneously only close to the wave source or to a defect that leads to mode conversion. By making the comparison between the in-plane and out-of-plane wave vector components the detection of mode conversion is possible, allowing for superior and reliable damage detection. Experimental verification of the proposed damage identification procedure is performed on fuel tank elements of Reusable Launch Vehicles designed for space exploration. Lamb waves are excited using low-profile, surface-bonded piezoceramic transducers and 3D scanning laser Doppler vibrometer is used to characterize the Lamb wave propagation field. The paper presents theoretical background of the proposed damage identification technique as well as experimental arrangements and results.

Fuel containment and damage tolerance for large composite primary aircraft structures. Phase 1: Testing

NASA Technical Reports Server (NTRS)

Sandifer, J. P.

1983-01-01

Technical problems associated with fuel containment and damage tolerance of composite material wings for transport aircraft were identified. The major tasks are the following: (1) the preliminary design of damage tolerant wing surface using composite materials; (2) the evaluation of fuel sealing and lightning protection methods for a composite material wing; and (3) an experimental investigation of the damage tolerant characteristics of toughened resin graphite/epoxy materials. The test results, the test techniques, and the test data are presented.

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

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.

The life cycle assessment of alternative fuel chains for urban buses and trolleybuses.

PubMed

Kliucininkas, L; Matulevicius, J; Martuzevicius, D

2012-05-30

This paper describes a comparative analysis of public transport alternatives in the city of Kaunas, Lithuania. An LCA (Life Cycle Assessment) inventory analysis of fuel chains was undertaken using the midi urban bus and a similar type of trolleybus. The inventory analysis of fuel chains followed the guidelines provided by the ISO 14040 and ISO 14044 standards. The ReCiPe Life Cycle Impact Assessment (LCIA) methodology was used to quantify weighted damage originating from five alternative fuel chains. The compressed biogas fuel chain had the lowest weighted damage value, namely 45.7 mPt/km, whereas weighted damage values of the fuel chains based on electricity generation for trolleybuses were 60.6 mPt/km (for natural gas) and 78.9 mPt/km (for heavy fuel oil). The diesel and compressed natural gas fuel chains exhibited considerably higher damage values of 114.2 mPt/km and 132.6 mPt/km, respectively. The comparative life cycle assessment of fuel chains suggested that biogas-powered buses and electric trolleybuses can be considered as the best alternatives to use when modernizing the public transport fleet in Kaunas. Copyright © 2012 Elsevier Ltd. All rights reserved.

Statistical Models of Fracture Relevant to Nuclear-Grade Graphite: Review and Recommendations

NASA Technical Reports Server (NTRS)

Nemeth, Noel N.; Bratton, Robert L.

2011-01-01

The nuclear-grade (low-impurity) graphite needed for the fuel element and moderator material for next-generation (Gen IV) reactors displays large scatter in strength and a nonlinear stress-strain response from damage accumulation. This response can be characterized as quasi-brittle. In this expanded review, relevant statistical failure models for various brittle and quasi-brittle material systems are discussed with regard to strength distribution, size effect, multiaxial strength, and damage accumulation. This includes descriptions of the Weibull, Batdorf, and Burchell models as well as models that describe the strength response of composite materials, which involves distributed damage. Results from lattice simulations are included for a physics-based description of material breakdown. Consideration is given to the predicted transition between brittle and quasi-brittle damage behavior versus the density of damage (level of disorder) within the material system. The literature indicates that weakest-link-based failure modeling approaches appear to be reasonably robust in that they can be applied to materials that display distributed damage, provided that the level of disorder in the material is not too large. The Weibull distribution is argued to be the most appropriate statistical distribution to model the stochastic-strength response of graphite.

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.

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.

78 FR 68688 - Airworthiness Directives; EADS CASA (Type Certificate Previously Held by Construcciones...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-15

... fuel transfer system. We are issuing this AD to detect and correct damage to certain fuel booster pumps... problem with the fuel transfer system. The results of the subsequent investigation revealed damage on the... was prompted by a report of an in-flight problem with the fuel transfer system. We are issuing this AD...

Automatic voltage imbalance detector

DOEpatents

Bobbett, Ronald E.; McCormick, J. Byron; Kerwin, William J.

1984-01-01

A device for indicating and preventing damage to voltage cells such as galvanic cells and fuel cells connected in series by detecting sequential voltages and comparing these voltages to adjacent voltage cells. The device is implemented by using operational amplifiers and switching circuitry is provided by transistors. The device can be utilized in battery powered electric vehicles to prevent galvanic cell damage and also in series connected fuel cells to prevent fuel cell damage.

Advances in structural damage assessment using strain measurements and invariant shape descriptors

NASA Astrophysics Data System (ADS)

Patki, Amol Suhas

Energy conservation has become one of the most important topic of engineering research over the last couple of decades all around the world and implies reduced energy consumption in order to preserve rapidly depleting natural resources. Along with development of fuel-efficient power plants and technology utilizing alternate fuel to traditional fossil fuels, the design and manufacturing of light-weight energy-efficient structures plays a major role in energy conservation. However this reduction in material and/or weight cannot be achieved at the expense of safety. Thus it is essential to either increase the confidence in the analysis of mechanics of traditional isotropic materials to reduce safety factors or develop new structural materials, such as fiber-reinforced (FRP) polymer matrix composites, which tend to have a higher strength to weight ratio. This doctoral research work will focus on two problems faced by the structural mechanics community viz. effects of closure and overloads on fatigue cracks and structural health monitoring of composites. Fatigue life prediction is largely empirical which in recent years has been shown to be a conservative design model. Investigation of crack growth mechanisms, such as crack closure can lead to design optimization. However, the lack of understanding and accepted theories introduces a degree of uncertainty in such models. Many of the complexity and uncertainty arise from the lack of an experimental technique to quantify crack closure. In this context, this research work offers the most compelling evidence to date of the effects of overload retardation and a confirmation of the Wheeler model using direct experimental observations of the stress field and crack tip plastic zone with the aid of thermoelastic stress analysis. On the other hand, the uncertainties in the post-damage behavior of energy saving FRP-composite materials increase their capital cost and maintenance cost. Damage in isotropic materials tends to be local to the area surrounding the damage, while damage in orthotropic materials tends to have more global repercussions. This calls for analysis of full-field strain distributions adding to the complexity of post-damage life estimation. This study explores shape descriptors used in the field of medical imagery, military targeting and biometric recognition for obtaining a qualitative and quantitative comparison between full-field strain data recorded from damaged composite panels using sophisticated experimental techniques. These descriptors are capable of decomposing images with 103 to 106 pixels into a feature vector with only a few hundred elements. This ability of shape descriptors to achieve enormous reduction in strain data, while providing unique representation, makes them a practical choice for the purpose of structural damage assessment. Consequently, it is relatively easy to statistically compare the shape descriptors of the full-field strain maps using similarity measures rather than the strain maps themselves. However, the wide range of geometric and design features in engineering components pose difficulties in the application of traditional shape description techniques. Thus a new shape descriptor is developed which is applicable to a wide range of specimen geometries. This work also illustrates how shape description techniques can be applied to full-field finite element model validations and updating.

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.

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...

Testing and Analysis of a Composite Non-Cylindrical Aircraft Fuselage Structure . Part II; Severe Damage

NASA Technical Reports Server (NTRS)

Przekop, Adam; Jegley, Dawn C.; Lovejoy, Andrew E.; Rouse, Marshall; Wu, Hsi-Yung T.

2016-01-01

The Environmentally Responsible Aviation Project aimed to develop aircraft technologies enabling significant fuel burn and community noise reductions. Small incremental changes to the conventional metallic alloy-based 'tube and wing' configuration were not sufficient to achieve the desired metrics. One airframe concept identified by the project as having the potential to dramatically improve aircraft performance was a composite-based hybrid wing body configuration. Such a concept, however, presented inherent challenges stemming from, among other factors, the necessity to transfer wing loads through the entire center fuselage section which accommodates a pressurized cabin confined by flat or nearly flat panels. This paper discusses a finite element analysis and the testing of a large-scale hybrid wing body center section structure developed and constructed to demonstrate that the Pultruded Rod Stitched Efficient Unitized Structure concept can meet these challenging demands of the next generation airframes. Part II of the paper considers the final test to failure of the test article in the presence of an intentionally inflicted severe discrete source damage under the wing up-bending loading condition. Finite element analysis results are compared with measurements acquired during the test and demonstrate that the hybrid wing body test article was able to redistribute and support the required design loads in a severely damaged condition.

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.

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.

Stability Analysis of Intertank Formed Skin/Stringer Compression Panel with Simulated Damage

NASA Technical Reports Server (NTRS)

Harper, David W.; Wingate, Robert J.

2012-01-01

The External Tank (ET) is a component of the Space Shuttle launch vehicle that contains fuel and oxidizer. During launch, the ET supplies the space shuttle main engines with liquid hydrogen and liquid oxygen. In addition to supplying fuel and oxidizer, it is the backbone structural component of the Space Shuttle. It is comprised of a liquid hydrogen (LH2) tank and a liquid oxygen (LOX) tank, which are separated by an Intertank. The Intertank is a stringer-stiffened cylindrical structure with hat-section stringers that are roll formed from aluminum-lithium alloy Al-2090. Cracks in the Intertank stringers of the STS-133 ET were noticed after a November 5, 2010 launch attempt. The cracks were approximately nine inches long and occurred on the forward end of the Intertank (near the LOX tank), along the fastener line, and were believed to have occurred while loading the ET with the cryogenic propellants. These cracks generated questions about the structural integrity of the Intertank. In order to determine the structural capability of the Intertank with varying degrees of damage, a finite element model (FEM) simulating a 1995 compression panel test was analyzed and correlated to test data. Varying degrees of damage were simulated in the FEM, and non-linear stability analyses were performed. The high degree of similarity between the compression panel and the Intertank provided confidence that the ET Intertank would have similar capabilities.

Uranium Pyrophoricity Phenomena and Prediction (FAI/00-39)

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

PLYS, M.G.

2000-10-10

The purpose of this report is to provide a topical reference on the phenomena and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel (SNF) Project with specific applications to SNF Project processes and situations. Spent metallic uranium nuclear fuel is currently stored underwater at the K basins in the Hanford 100 area, and planned processing steps include: (1) At the basins, cleaning and placing fuel elements and scrap into stainless steel multi-canister overpacks (MCOs) holding about 6 MT of fuel apiece; (2) At nearby cold vacuum drying (CVD) stations, draining, vacuum drying, and mechanically sealing the MCOs; (3)more » Shipping the MCOs to the Canister Storage Building (CSB) on the 200 Area plateau; and (4) Welding shut and placing the MCOs for interim (40 year) dry storage in closed CSB storage tubes cooled by natural air circulation through the surrounding vault. Damaged fuel elements have exposed and corroded fuel surfaces, which can exothermically react with water vapor and oxygen during normal process steps and in off-normal situations, A key process safety concern is the rate of reaction of damaged fuel and the potential for self-sustaining or runaway reactions, also known as uranium fires or fuel ignition. Uranium metal and one of its corrosion products, uranium hydride, are potentially pyrophoric materials. Dangers of pyrophoricity of uranium and its hydride have long been known in the U.S. Department of Energy (Atomic Energy Commission/DOE) complex and will be discussed more below; it is sufficient here to note that there are numerous documented instances of uranium fires during normal operations. The motivation for this work is to place the safety of the present process in proper perspective given past operational experience. Steps in development of such a perspective are: (1) Description of underlying physical causes for runaway reactions, (2) Modeling physical processes to explain runaway reactions, (3) Validation of the method against experimental data, (4) Application of the method to plausibly explain operational experience, and (5) Application of the method to present process steps to demonstrate process safety and margin. Essentially, the logic above is used to demonstrate that runaway reactions cannot occur during normal SNF Project process steps, and to illustrate the depth of the technical basis for such a conclusion. Some off-normal conditions are identified here that could potentially lead to runaway reactions. However, this document is not intended to provide an exhaustive analysis of such cases. In summary, this report provides a ''toolkit'' of models and approaches for analysis of pyrophoricity safety issues at Hanford, and the technical basis for the recommended approaches. A summary of recommended methods appears in Section 9.0.« less

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)

Investigation of thermoelastic stresses induced at high altitudes on aircraft external fuel tanks

NASA Astrophysics Data System (ADS)

Mousseau, Stephanie Lynn Steber

As composite technology has grown over the past several decades, the use of composite materials in military applications has become more feasible and widely accepted. Although composite materials provide many benefits, including strength optimization and reduced weight, damage and repair of these materials creates an additional challenge, especially when operating in a marine environment, such as on a carrier deck. This is evident within the Navy, as excessive damage often leads to the scrapping of F/A-18 External Fuel Tanks. This damage comes in many forms, the most elusive of which is delamination. Often the delamination found on the tanks is beyond repairable limits and the cause unknown, making it difficult to predict and prevent. The purpose of this investigation was to study the structure of the Navy's 330 gallon External Fuel Tanks and investigate one potential cause of delamination, stresses induced at high altitudes by cold temperatures. A stress analysis was completed using finite element software, and validation of the model was accomplished through testing of a scale model specimen. Due to the difficulties in modeling and predicting delamination, such as unknown presence of voids and understanding failure criteria, delamination was not modeled in Abaqus, rather stresses were observed and characteristics were studied to understand the potential for delamination within the layup. In addition, studies were performed to understand the effect of material properties and layup sequence on the stress distribution within the tank. Alternative design solutions are presented which could reduce the radial stresses within the tank, and recommendations are made for further study to understand the trade-offs between stress, cost, and manufacturability.

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.

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

Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

DTIC Science & Technology

1990-06-01

long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

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...

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.

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.

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)

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

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

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish releasemore » fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.« less

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

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.

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)

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

NASA Astrophysics Data System (ADS)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

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.

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

14 CFR 27.977 - Fuel tank outlet.

Code of Federal Regulations, 2012 CFR

2012-01-01

... STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.977 Fuel tank outlet. (a) There must be a... rotorcraft, prevent the passage of any object that could restrict fuel flow or damage any fuel system...] Fuel System Components ...

14 CFR 27.977 - Fuel tank outlet.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.977 Fuel tank outlet. (a) There must be a... rotorcraft, prevent the passage of any object that could restrict fuel flow or damage any fuel system...] Fuel System Components ...

14 CFR 27.977 - Fuel tank outlet.

Code of Federal Regulations, 2014 CFR

2014-01-01

... STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.977 Fuel tank outlet. (a) There must be a... rotorcraft, prevent the passage of any object that could restrict fuel flow or damage any fuel system...] Fuel System Components ...

14 CFR 27.977 - Fuel tank outlet.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.977 Fuel tank outlet. (a) There must be a... rotorcraft, prevent the passage of any object that could restrict fuel flow or damage any fuel system...] Fuel System Components ...

14 CFR 27.977 - Fuel tank outlet.

Code of Federal Regulations, 2013 CFR

2013-01-01

... STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.977 Fuel tank outlet. (a) There must be a... rotorcraft, prevent the passage of any object that could restrict fuel flow or damage any fuel system...] Fuel System Components ...

Advances in threat assessment and their application to forest and rangeland management—Volume 2

Treesearch

H. Michael Rauscher; Yasmeen Sands; Danny C. Lee; Jerome S. Beatty

2010-01-01

Risk is a combined statement of the probability that something of value will be damaged and some measure of the damage’s adverse effect. Wildfires burning in the uncharacteristic fuel conditions now typical throughout the Western United States can damage ecosystems and adversely affect environmental conditions. Wildfire behavior can be modified by prefire fuel...

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

Han, Bo-Young; Choi, Daewoong; Park, Se Hwan

Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in themore » target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)« less

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)

Fuel containment, lightning protection and damage tolerance in large composite primary aircraft structures

NASA Technical Reports Server (NTRS)

Griffin, Charles F.; James, Arthur M.

1985-01-01

The damage-tolerance characteristics of high strain-to-failure graphite fibers and toughened resins were evaluated. Test results show that conventional fuel tank sealing techniques are applicable to composite structures. Techniques were developed to prevent fuel leaks due to low-energy impact damage. For wing panels subjected to swept stroke lightning strikes, a surface protection of graphite/aluminum wire fabric and a fastener treatment proved effective in eliminating internal sparking and reducing structural damage. The technology features developed were incorporated and demonstrated in a test panel designed to meet the strength, stiffness, and damage tolerance requirements of a large commercial transport aircraft. The panel test results exceeded design requirements for all test conditions. Wing surfaces constructed with composites offer large weight savings if design allowable strains for compression can be increased from current levels.

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.

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.

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.

Abstracts: Energy Sciences programs, January--December 1978

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

None

This report presents abstracts of all publications in the Energy Sciences programs of the Department of Energy and Environment from January 1, 1978 through December 31, 1978. It is a companion report to Annual Highlights of Programs in Energy Sciences - (December 1978, BNL 50973). Together, they present scientific and/or technical highlights of the Energy Sciences programs for the past calendar year, detailed descriptions of all the programs, and the publication issuing from the work performed. The following are some of the topics included: porphyrin chemistry; chemistry of energetic compounds; combustion; coal utilization; metal hydrides; cyclic separations process research; tracemore » element analysis; materials properties and structures; radiation damage; superconducting materials; materials of construction for geothermal applications; repair of deteriorated concrete; development of glass--polymer composite sewer pipe; flash hydropyrolysis of coal; desulfurization of high-temperature combustion and fuel gases; and synthetic fuels development. (RWR)« less

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.)

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)

Triaxial Swirl Injector Element for Liquid-Fueled Engines

NASA Technical Reports Server (NTRS)

Muss, Jeff

2010-01-01

A triaxial injector is a single bi-propellant injection element located at the center of the injector body. The injector element consists of three nested, hydraulic swirl injectors. A small portion of the total fuel is injected through the central hydraulic injector, all of the oxidizer is injected through the middle concentric hydraulic swirl injector, and the balance of the fuel is injected through an outer concentric injection system. The configuration has been shown to provide good flame stabilization and the desired fuel-rich wall boundary condition. The injector design is well suited for preburner applications. Preburner injectors operate at extreme oxygen-to-fuel mass ratios, either very rich or very lean. The goal of a preburner is to create a uniform drive gas for the turbomachinery, while carefully controlling the temperature so as not to stress or damage turbine blades. The triaxial injector concept permits the lean propellant to be sandwiched between two layers of the rich propellant, while the hydraulic atomization characteristics of the swirl injectors promote interpropellant mixing and, ultimately, good combustion efficiency. This innovation is suited to a wide range of liquid oxidizer and liquid fuels, including hydrogen, methane, and kerosene. Prototype testing with the triaxial swirl injector demonstrated excellent injector and combustion chamber thermal compatibility and good combustion performance, both at levels far superior to a pintle injector. Initial testing with the prototype injector demonstrated over 96-percent combustion efficiency. The design showed excellent high -frequency combustion stability characteristics with oxygen and kerosene propellants. Unlike the more conventional pintle injector, there is not a large bluff body that must be cooled. The absence of a protruding center body enhances the thermal durability of the triaxial swirl injector. The hydraulic atomization characteristics of the innovation allow the design to be rapidly scaled from small in-space applications [500-5,000 lbf (2.2 22.2 kN)] to large thrust engine applications [80,000 lbf (356 kN) and beyond]. The triaxial injector is also less sensitive to eccentricities, manufacturing tolerances, and gap width of many traditional coaxial and pintle injector designs. The triaxial-injector injection orifice configuration provides for high injection stiffness. The low parts count and relatively large injector design features are amenable to low-cost production.

Design of a fuel element for a lead-cooled fast reactor

NASA Astrophysics Data System (ADS)

Sobolev, V.; Malambu, E.; Abderrahim, H. Aït

2009-03-01

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg-1 of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg-1 of HM.

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.

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

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...

75 FR 34347 - Airworthiness Directives; Honeywell International Inc. Auxiliary Power Unit Models GTCP36-150(R...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-17

... lead to ignition of fuel vapor, creating a fire and explosion hazard resulting in injury, and damage to... ignition of fuel vapor, creating a fire and explosion hazard resulting in injury, and damage to the APU and...

76 FR 12825 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Confirmation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-09

... definitions for Damaged Fuel Assembly and Transfer Operations; add definitions for Fuel Class and Reconstituted Fuel Assembly; add Combustion Engineering 16x16 class fuel assemblies as authorized contents...

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

Methods and Piezoelectric Imbedded Sensors for Damage Detection in Composite Plates Under Ambient and Cryogenic Conditions

NASA Technical Reports Server (NTRS)

Engberg, Robert; Ooi, Teng K.

2004-01-01

New methods for structural health monitoring are being assessed, especially in high-performance, extreme environment, safety-critical applications. One such application is for composite cryogenic fuel tanks. The work presented here attempts to characterize and investigate the feasibility of using imbedded piezoelectric sensors to detect cracks and delaminations under cryogenic and ambient conditions. A variety of damage detection methods and different Sensors are employed in the different composite plate samples to aid in determining an optimal algorithm, sensor placement strategy, and type of imbedded sensor to use. Variations of frequency, impedance measurements, and pulse echoing techniques of the sensors are employed and compared. Statistical and analytic techniques are then used to determine which method is most desirable for a specific type of damage. These results are furthermore compared with previous work using externally mounted sensors. Results and optimized methods from this work can then be incorporated into a larger composite structure to validate and assess its structural health. This could prove to be important in the development and qualification of any 2" generation reusable launch vehicle using composites as a structural element.

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.

Assessment of infrastructure functional damages caused by natural-technological disasters

NASA Astrophysics Data System (ADS)

Massabò, Marco; Trasforini, Eva; Traverso, Stefania; Rudari, Roberto; De Angeli, Silvia; Cecinati, Francesca; Cerruti, Valentina

2013-04-01

The assessment of infrastructure damages caused by technological disaster poses several challenges, from gathering needed information on the territorial system to the definition of functionality curves for infrastructures elements (such as, buildings, road school) that are exposed to both natural and technological event. Moreover, areas affected by natural or natech (technological disasters triggered by natural events) disasters have often very large extensions and a rapid survey of them to gather all the needed information is a very difficult task, for many reasons, not least the difficult access to the existing databases and resources. We use multispectral optical imagery with other geographical and unconventional data to identify and characterize exposed elements. Our efforts in the virtual survey and during the investigation steps have different aims: to identify the vulnerability of infrastructures, buildings or activities; to execute calculations of exposition to risk; to estimate physical and functional damages. Subsequently, we apply specific algorithms to estimate values of acting forces and physical and functional damages. The updated picture of target areas in terms of risk-prone people, infrastructures and their connections is very important. It is possible to develop algorithms providing values of systemic functionality for each network element. The methodology is here applied to a natech disaster, arising from the combination of a flood event (specifically, the January 2010 flooding of Drin and Buna rivers, with a worsening in the road safety levels in the Shkoder area) with and the subsequent overturning of a truck transporting hazardous material. The accident causes the loss of containment and the total material release. Once the release has taken place, the evolution will depend on the physical state of the substance spilled (liquid, gas or dust). As a specific case we consider the rupture of a trucks transporting liquid fuels such as gasoline through Shkoder downtown. Goods entering in Albania from north pass through Shkoder, indeed a high traffic road that connects Albania with Montenegro and Kosovo crosses Shkoder downtown. We consider a truck overturned in downtown Shkoder during the flooding of January 2010; the gasoline transported by the truck is completely released and a pool fire develops damaging roads. We use the model CHESRM (Chemical Spill Risk Mapper) for identify the threat zones of the accident and as a basis for assessing the potential leads to functional damages to other elements of the considered system. The application of the methodology shows the potential use not only on real time emergency management or prevention but also during post-event management for the evaluation of the functional damage to the affected infrastructure (villages isolated from the rest of the network, villages unable to reach schools, hospitals or other services...) and to set a hierarchy in restoration activities, giving priority to the reconstruction of links between primary nodes.

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.

Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

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

Ibarra, Luis; Medina, Ricardo; Yang, Haori

This study assessed the risk of loss of structural integrity of transportation casks and fuel cladding after extended storage. Although it is known that fuel rods discharged from NPPs have a small percentage of rod cladding defects, the behavior of fuel cladding and the structural elements of assemblies during transportation after long-term storage is not well understood. If the fuel degrades during extended storage, it could be susceptible to damage from vibration and impact loads during transport operations, releasing fission-product gases into the canister or the cask interior (NWTRB 2010). Degradation of cladding may occur due to mechanisms associated withmore » hydrogen embrittlement, delayed hydride cracking, low temperature creep, and stress corrosion cracking (SCC) that may affect fuel cladding and canister components after extended storage of hundreds of years. Over extended periods at low temperatures, these mechanisms affect the ductility, strength, and fracture toughness of the fuel cladding, which becomes brittle. For transportation purposes, the fuel may be transferred from storage to shipping casks, or dual-purpose casks may be used for storage and transportation. Currently, most of the transportation casks will be the former case. A risk assessment evaluation is conducted based on results from experimental tests and simulations with advanced numerical models. A novel contribution of this study is the evaluation of the combined effect of component aging and vibration/impact loads in transportation scenarios. The expected levels of deterioration will be obtained from previous and current studies on the effect of aging on fuel and cask components. The emphasis of the study is placed on the structural integrity of fuel cladding and canisters.« less

76 FR 18960 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-06

... unsafe condition is damage to wiring in the wing, center, and trim fuel tanks, due to failed P-clips used..., center, or trim fuel tanks. The proposed AD would require actions that are intended to address the unsafe..., 2009]. The unsafe condition is damage to wiring in the wing, center, and trim fuel tanks, due to failed...

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.

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.

KSC-2010-4748

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- The Space Shuttle Program's last external fuel tank, ET-122, is loaded onto the Pegasus Barge at NASA's Michoud Assembly Facility in New Orleans. The tank will travel 900 miles to NASA's Kennedy Space Center in Florida where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett

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.

Method to reduce damage to backing plate

DOEpatents

Perry, Michael D.; Banks, Paul S.; Stuart, Brent C.

2001-01-01

The present invention is a method for penetrating a workpiece using an ultra-short pulse laser beam without causing damage to subsequent surfaces facing the laser. Several embodiments are shown which place holes in fuel injectors without damaging the back surface of the sack in which the fuel is ejected. In one embodiment, pulses from an ultra short pulse laser remove about 10 nm to 1000 nm of material per pulse. In one embodiment, a plasma source is attached to the fuel injector and initiated by common methods such as microwave energy. In another embodiment of the invention, the sack void is filled with a solid. In one other embodiment, a high viscosity liquid is placed within the sack. In general, high-viscosity liquids preferably used in this invention should have a high damage threshold and have a diffusing property.

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.

Space shuttle orbit maneuvering engine, reusable thrust chamber program. Task 6: Data dump hot fuel element investigation

NASA Technical Reports Server (NTRS)

Nurick, W. H.

1974-01-01

An evaluation of reusable thrust chambers for the space shuttle orbit maneuvering engine was conducted. Tests were conducted using subscale injector hot-fire procedures for the injector configurations designed for a regenerative cooled engine. The effect of operating conditions and fuel temperature on combustion chamber performance was determined. Specific objectives of the evaluation were to examine the optimum like-doublet element geometry for operation at conditions consistent with a fuel regeneratively cooled engine (hot fuel, 200 to 250 F) and the sensitivity of the triplet injector element to hot fuels.

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.

Modelling the side impact of carbon fibre tubes

NASA Astrophysics Data System (ADS)

Sudharsan, Ms R.; Rolfe, B. F., Dr; Hodgson, P. D., Prof

2010-06-01

Metallic tubes have been extensively studied for their crashworthiness as they closely resemble automotive crash rails. Recently, the demand to improve fuel economy and reduce vehicle emissions has led automobile manufacturers to explore the crash properties of light weight materials such as fibre reinforced polymer composites, metallic foams and sandwich structures in order to use them as crash barriers. This paper discusses the response of carbon fibre reinforced polymer (CFRP) tubes and their failure mechanisms during side impact. The energy absorption of CFRP tubes is compared to similar Aluminium tubes. The response of the CFRP tubes during impact was modelled using Abaqus finite element software with a composite fabric material model. The material inputs were given based on standard tension and compression test results and the in-plane damage was defined based on cyclic shear tests. The failure modes and energy absorption observed during the tests were well represented by the finite element model.

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.

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.

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.

14 CFR 29.977 - Fuel tank outlet.

Code of Federal Regulations, 2014 CFR

2014-01-01

... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Fuel tank outlet. 29.977 Section 29.977...

14 CFR 25.977 - Fuel tank outlet.

Code of Federal Regulations, 2014 CFR

2014-01-01

... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Fuel tank outlet. 25.977 Section 25.977...

14 CFR 23.977 - Fuel tank outlet.

Code of Federal Regulations, 2013 CFR

2013-01-01

... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.977 Fuel... damage any fuel system component. (b) The clear area of each fuel tank outlet strainer must be at least... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Fuel tank outlet. 23.977 Section 23.977...

14 CFR 23.977 - Fuel tank outlet.

Code of Federal Regulations, 2014 CFR

2014-01-01

... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.977 Fuel... damage any fuel system component. (b) The clear area of each fuel tank outlet strainer must be at least... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Fuel tank outlet. 23.977 Section 23.977...

14 CFR 29.977 - Fuel tank outlet.

Code of Federal Regulations, 2013 CFR

2013-01-01

... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Fuel tank outlet. 29.977 Section 29.977...

14 CFR 25.977 - Fuel tank outlet.

Code of Federal Regulations, 2013 CFR

2013-01-01

... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Fuel tank outlet. 25.977 Section 25.977...

14 CFR 25.977 - Fuel tank outlet.

Code of Federal Regulations, 2012 CFR

2012-01-01

... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2012-01-01 2012-01-01 false Fuel tank outlet. 25.977 Section 25.977...

14 CFR 29.977 - Fuel tank outlet.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Fuel tank outlet. 29.977 Section 29.977...

14 CFR 29.977 - Fuel tank outlet.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank outlet. 29.977 Section 29.977...

14 CFR 25.977 - Fuel tank outlet.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank outlet. 25.977 Section 25.977...

14 CFR 23.977 - Fuel tank outlet.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.977 Fuel... damage any fuel system component. (b) The clear area of each fuel tank outlet strainer must be at least... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank outlet. 23.977 Section 23.977...

14 CFR 25.977 - Fuel tank outlet.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Fuel tank outlet. 25.977 Section 25.977...

14 CFR 23.977 - Fuel tank outlet.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.977 Fuel... damage any fuel system component. (b) The clear area of each fuel tank outlet strainer must be at least... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Fuel tank outlet. 23.977 Section 23.977...

14 CFR 23.977 - Fuel tank outlet.

Code of Federal Regulations, 2012 CFR

2012-01-01

... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.977 Fuel... damage any fuel system component. (b) The clear area of each fuel tank outlet strainer must be at least... 14 Aeronautics and Space 1 2012-01-01 2012-01-01 false Fuel tank outlet. 23.977 Section 23.977...

14 CFR 29.977 - Fuel tank outlet.

Code of Federal Regulations, 2012 CFR

2012-01-01

... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.977 Fuel tank outlet. (a) There must be... airplanes, prevent the passage of any object that could restrict fuel flow or damage any fuel system... 14 Aeronautics and Space 1 2012-01-01 2012-01-01 false Fuel tank outlet. 29.977 Section 29.977...

An experimental demonstration of stem damage as a predictor of fire-caused mortality for ponderosa pine

USGS Publications Warehouse

van Mantgem, P.; Schwartz, M.

2004-01-01

We subjected 159 small ponderosa pine (Pinus ponderosa Dougl. ex P. & C. Laws.) to treatments designed to test the relative importance of stem damage as a predictor of postfire mortality. The treatments consisted of a group with the basal bark artificially thinned, a second group with fuels removed from the base of the stem, and an untreated control. Following prescribed burning, crown scorch severity was equivalent among the groups. Postfire mortality was significantly less frequent in the fuels removal group than in the bark removal and control groups. No model of mortality for the fuels removal group was possible, because dead trees constituted <4% of subject trees. Mortality in the bark removal group was best predicted by crown scorch and stem scorch severity, whereas death in the control group was predicted by crown scorch severity and bark thickness. The relative lack of mortality in the fuels removal group and the increased sensitivity to stem damage in the bark removal group suggest that stem damage is a critical determinant of postfire mortality for small ponderosa pine.

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/.

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.

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.

14 CFR 25.994 - Fuel system components.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

14 CFR 25.994 - Fuel system components.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

14 CFR 25.994 - Fuel system components.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 14 Aeronautics and Space 1 2012-01-01 2012-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

14 CFR 25.994 - Fuel system components.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

14 CFR 25.994 - Fuel system components.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

Prescribed burning in ponderosa pine: fuel reductions and redistributing fuels near boles to prevent injury

USDA-ARS?s Scientific Manuscript database

Prescribed burning can be an effective tool for thinning forests and reducing fuels to lessen wildfire risks. However, prescribed burning sometimes fails to substantially reduce fuels and sometimes damages/kills valuable, large trees. This study compared fuel reductions between fall and spring pre...

Computational predictive methods for fracture and fatigue

NASA Technical Reports Server (NTRS)

Cordes, J.; Chang, A. T.; Nelson, N.; Kim, Y.

1994-01-01

The damage-tolerant design philosophy as used by aircraft industries enables aircraft components and aircraft structures to operate safely with minor damage, small cracks, and flaws. Maintenance and inspection procedures insure that damages developed during service remain below design values. When damage is found, repairs or design modifications are implemented and flight is resumed. Design and redesign guidelines, such as military specifications MIL-A-83444, have successfully reduced the incidence of damage and cracks. However, fatigue cracks continue to appear in aircraft well before the design life has expired. The F16 airplane, for instance, developed small cracks in the engine mount, wing support, bulk heads, the fuselage upper skin, the fuel shelf joints, and along the upper wings. Some cracks were found after 600 hours of the 8000 hour design service life and design modifications were required. Tests on the F16 plane showed that the design loading conditions were close to the predicted loading conditions. Improvements to analytic methods for predicting fatigue crack growth adjacent to holes, when multiple damage sites are present, and in corrosive environments would result in more cost-effective designs, fewer repairs, and fewer redesigns. The overall objective of the research described in this paper is to develop, verify, and extend the computational efficiency of analysis procedures necessary for damage tolerant design. This paper describes an elastic/plastic fracture method and an associated fatigue analysis method for damage tolerant design. Both methods are unique in that material parameters such as fracture toughness, R-curve data, and fatigue constants are not required. The methods are implemented with a general-purpose finite element package. Several proof-of-concept examples are given. With further development, the methods could be extended for analysis of multi-site damage, creep-fatigue, and corrosion fatigue problems.

Computational predictive methods for fracture and fatigue

NASA Astrophysics Data System (ADS)

Cordes, J.; Chang, A. T.; Nelson, N.; Kim, Y.

1994-09-01

The damage-tolerant design philosophy as used by aircraft industries enables aircraft components and aircraft structures to operate safely with minor damage, small cracks, and flaws. Maintenance and inspection procedures insure that damages developed during service remain below design values. When damage is found, repairs or design modifications are implemented and flight is resumed. Design and redesign guidelines, such as military specifications MIL-A-83444, have successfully reduced the incidence of damage and cracks. However, fatigue cracks continue to appear in aircraft well before the design life has expired. The F16 airplane, for instance, developed small cracks in the engine mount, wing support, bulk heads, the fuselage upper skin, the fuel shelf joints, and along the upper wings. Some cracks were found after 600 hours of the 8000 hour design service life and design modifications were required. Tests on the F16 plane showed that the design loading conditions were close to the predicted loading conditions. Improvements to analytic methods for predicting fatigue crack growth adjacent to holes, when multiple damage sites are present, and in corrosive environments would result in more cost-effective designs, fewer repairs, and fewer redesigns. The overall objective of the research described in this paper is to develop, verify, and extend the computational efficiency of analysis procedures necessary for damage tolerant design. This paper describes an elastic/plastic fracture method and an associated fatigue analysis method for damage tolerant design. Both methods are unique in that material parameters such as fracture toughness, R-curve data, and fatigue constants are not required. The methods are implemented with a general-purpose finite element package. Several proof-of-concept examples are given. With further development, the methods could be extended for analysis of multi-site damage, creep-fatigue, and corrosion fatigue problems.

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.

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.

Comparison of burning characteristics of live and dead chaparral fuels

Treesearch

L. Sun; X. Zhou; S. Mahalingam; D.R. Weise

2006-01-01

Wildfire spread in living vegetation, such as chaparral in southern California, often causes significant damage to infrastructure and ecosystems. The effects of physical characteristics of fuels and fuel beds on live fuel burning and whether live fuels differ fundamentally from dead woody fuels in their burning characteristics are not well understood. Toward this end,...

Simulating Impacts of Disruptions to Liquid Fuels Infrastructure

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

Wilson, Michael; Corbet, Thomas F.; Baker, Arnold B.

This report presents a methodology for estimating the impacts of events that damage or disrupt liquid fuels infrastructure. The impact of a disruption depends on which components of the infrastructure are damaged, the time required for repairs, and the position of the disrupted components in the fuels supply network. Impacts are estimated for seven stressing events in regions of the United States, which were selected to represent a range of disruption types. For most of these events the analysis is carried out using the National Transportation Fuels Model (NTFM) to simulate the system-level liquid fuels sector response. Results are presentedmore » for each event, and a brief cross comparison of event simulation results is provided.« less

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.

Detection of multiple damages employing best achievable eigenvectors under Bayesian inference

NASA Astrophysics Data System (ADS)

Prajapat, Kanta; Ray-Chaudhuri, Samit

2018-05-01

A novel approach is presented in this work to localize simultaneously multiple damaged elements in a structure along with the estimation of damage severity for each of the damaged elements. For detection of damaged elements, a best achievable eigenvector based formulation has been derived. To deal with noisy data, Bayesian inference is employed in the formulation wherein the likelihood of the Bayesian algorithm is formed on the basis of errors between the best achievable eigenvectors and the measured modes. In this approach, the most probable damage locations are evaluated under Bayesian inference by generating combinations of various possible damaged elements. Once damage locations are identified, damage severities are estimated using a Bayesian inference Markov chain Monte Carlo simulation. The efficiency of the proposed approach has been demonstrated by carrying out a numerical study involving a 12-story shear building. It has been found from this study that damage scenarios involving as low as 10% loss of stiffness in multiple elements are accurately determined (localized and severities quantified) even when 2% noise contaminated modal data are utilized. Further, this study introduces a term parameter impact (evaluated based on sensitivity of modal parameters towards structural parameters) to decide the suitability of selecting a particular mode, if some idea about the damaged elements are available. It has been demonstrated here that the accuracy and efficiency of the Bayesian quantification algorithm increases if damage localization is carried out a-priori. An experimental study involving a laboratory scale shear building and different stiffness modification scenarios shows that the proposed approach is efficient enough to localize the stories with stiffness modification.

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.

Default operational intervention levels (OILs) for severe nuclear power plant or spent fuel pool emergencies.

PubMed

McKenna, T; Kutkov, V; Vilar Welter, P; Dodd, B; Buglova, E

2013-05-01

Experience and studies show that for an emergency at a nuclear power plant involving severe core damage or damage to the fuel in spent fuel pools, the following actions may need to be taken in order to prevent severe deterministic health effects and reduce stochastic health effects: (1) precautionary protective actions and other response actions for those near the facility (i.e., within the zones identified by the International Atomic Energy Agency) taken immediately upon detection of facility conditions indicating possible severe damage to the fuel in the core or in the spent fuel pool; and (2) protective actions and other response actions taken based on environmental monitoring and sampling results following a release. This paper addresses the second item by providing default operational intervention levels [OILs, which are similar to the U.S. derived response levels (DRLs)] for promptly assessing radioactive material deposition, as well as skin, food, milk and drinking water contamination, following a major release of fission products from the core or spent fuel pool of a light water reactor (LWR) or a high power channel reactor (RBMK), based on the International Atomic Energy Agency's guidance.

A Coupled/Uncoupled Computational Scheme for Deformation and Fatigue Damage Analysis of Unidirectional Metal-Matrix Composites

NASA Technical Reports Server (NTRS)

Wilt, Thomas E.; Arnold, Steven M.; Saleeb, Atef F.

1997-01-01

A fatigue damage computational algorithm utilizing a multiaxial, isothermal, continuum-based fatigue damage model for unidirectional metal-matrix composites has been implemented into the commercial finite element code MARC using MARC user subroutines. Damage is introduced into the finite element solution through the concept of effective stress that fully couples the fatigue damage calculations with the finite element deformation solution. Two applications using the fatigue damage algorithm are presented. First, an axisymmetric stress analysis of a circumferentially reinforced ring, wherein both the matrix cladding and the composite core were assumed to behave elastic-perfectly plastic. Second, a micromechanics analysis of a fiber/matrix unit cell using both the finite element method and the generalized method of cells (GMC). Results are presented in the form of S-N curves and damage distribution plots.

75 FR 30274 - Airworthiness Directives; McDonnell Douglas Corporation Model MD-11 and MD-11F Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-01

... installed on wire harnesses of the tail tank fuel transfer pumps, and to determine if damaged wires are... a permanent modification of the wire harnesses if metallic transitions are not installed, which... harnesses of the tail tank fuel transfer pumps, and to determine if damaged wires are present; and repair...

Operation of the WWR-S Reactor in Poland and in the Period 1961-1962; DOKLAD OB ISPOL'ZOVANII I EKSPLUATATSII REAKTORA TIPA VVR-C V POL'SHE ZA 1961-1962 GODA

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

Aleksandrowicz, J.

1963-03-01

The experimental equipment used in the work at the horizontal reactor channels is listed. Diagrams of the utilization of the nominal reactor power and core loading are given, the reactivity fractions of the separate fuel assemblies are detonated, together with the diagram of reactivity versus burnup. Reactor channels and space used for sample irradiation and isotope production are described, and the total number of irradiations is given. Results of the measurements connected with the routine reactor operation are quoted, namely: analysis of water purity in the primary circuit, analysis of the work of the ion exchanger and mechanical filter, andmore » analysis of air activity in the special ventilation system. Data are given concerning radiation protection of personnel, including individual monitoring. Leak testing of the fuel elements is discussed. Damage of the reactor equipment and appearance of alarm signals are described. (auth)« less

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

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)

Guided wave propagation and spectral element method for debonding damage assessment in RC structures

NASA Astrophysics Data System (ADS)

Wang, Ying; Zhu, Xinqun; Hao, Hong; Ou, Jinping

2009-07-01

A concrete-steel interface spectral element is developed to study the guided wave propagation along the steel rebar in the concrete. Scalar damage parameters characterizing changes in the interface (debonding damage) are incorporated into the formulation of the spectral finite element that is used for damage detection of reinforced concrete structures. Experimental tests are carried out on a reinforced concrete beam with embedded piezoelectric elements to verify the performance of the proposed model and algorithm. Parametric studies are performed to evaluate the effect of different damage scenarios on wave propagation in the reinforced concrete structures. Numerical simulations and experimental results show that the method is effective to model wave propagation along the steel rebar in concrete and promising to detect damage in the concrete-steel interface.

A coupled/uncoupled deformation and fatigue damage algorithm utilizing the finite element method

NASA Technical Reports Server (NTRS)

Wilt, Thomas E.; Arnold, Steven M.

1994-01-01

A fatigue damage computational algorithm utilizing a multiaxial, isothermal, continuum based fatigue damage model for unidirectional metal matrix composites has been implemented into the commercial finite element code MARC using MARC user subroutines. Damage is introduced into the finite element solution through the concept of effective stress which fully couples the fatigue damage calculations with the finite element deformation solution. An axisymmetric stress analysis was performed on a circumferentially reinforced ring, wherein both the matrix cladding and the composite core were assumed to behave elastic-perfectly plastic. The composite core behavior was represented using Hill's anisotropic continuum based plasticity model, and similarly, the matrix cladding was represented by an isotropic plasticity model. Results are presented in the form of S-N curves and damage distribution plots.

Impacts of climate mitigation strategies in the energy sector on global land use and carbon balance

NASA Astrophysics Data System (ADS)

Engström, Kerstin; Lindeskog, Mats; Olin, Stefan; Hassler, John; Smith, Benjamin

2017-09-01

Reducing greenhouse gas emissions to limit damage to the global economy climate-change-induced and secure the livelihoods of future generations requires ambitious mitigation strategies. The introduction of a global carbon tax on fossil fuels is tested here as a mitigation strategy to reduce atmospheric CO2 concentrations and radiative forcing. Taxation of fossil fuels potentially leads to changed composition of energy sources, including a larger relative contribution from bioenergy. Further, the introduction of a mitigation strategy reduces climate-change-induced damage to the global economy, and thus can indirectly affect consumption patterns and investments in agricultural technologies and yield enhancement. Here we assess the implications of changes in bioenergy demand as well as the indirectly caused changes in consumption and crop yields for global and national cropland area and terrestrial biosphere carbon balance. We apply a novel integrated assessment modelling framework, combining three previously published models (a climate-economy model, a socio-economic land use model and an ecosystem model). We develop reference and mitigation scenarios based on the narratives and key elements of the shared socio-economic pathways (SSPs). Taking emissions from the land use sector into account, we find that the introduction of a global carbon tax on the fossil fuel sector is an effective mitigation strategy only for scenarios with low population development and strong sustainability criteria (SSP1 Taking the green road). For scenarios with high population growth, low technological development and bioenergy production the high demand for cropland causes the terrestrial biosphere to switch from being a carbon sink to a source by the end of the 21st century.

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

NASA Astrophysics Data System (ADS)

Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

2013-10-01

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

Artificial Boundary Conditions for Finite Element Model Update and Damage Detection

DTIC Science & Technology

2017-03-01

BOUNDARY CONDITIONS FOR FINITE ELEMENT MODEL UPDATE AND DAMAGE DETECTION by Emmanouil Damanakis March 2017 Thesis Advisor: Joshua H. Gordis...REPORT TYPE AND DATES COVERED Master’s thesis 4. TITLE AND SUBTITLE ARTIFICIAL BOUNDARY CONDITIONS FOR FINITE ELEMENT MODEL UPDATE AND DAMAGE DETECTION...release. Distribution is unlimited. 12b. DISTRIBUTION CODE 13. ABSTRACT (maximum 200 words) In structural engineering, a finite element model is often

KSC-2010-4747

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers escort the Space Shuttle Program's last external fuel tank, ET-122, to the Pegasus Barge at NASA's Michoud Assembly Facility in New Orleans. The tank will travel 900 miles aboard the Pegasus Barge to NASA's Kennedy Space Center in Florida where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett

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.

The effect of high energy ion beam analysis on D trapping in W

NASA Astrophysics Data System (ADS)

Finlay, T. J.; Davis, J. W.; Schwarz-Selinger, T.; Haasz, A. A.

2017-12-01

High energy ion beam analyses (IBA) are invaluable for measuring concentration depth profiles of light elements in solid materials, and important in the study of fusion fuel retention in tokamaks. Polycrystalline W specimens were implanted at 300 and 500 K, 5-10 × 1023 D m-2 fluence, with deuterium-only and simultaneous D-3%He ion beams. Selected specimens were analysed by elastic recoil detection analysis (ERDA) and/or nuclear reaction analysis (NRA). All specimens were measured by thermal desorption spectroscopy (TDS). The D TDS spectra show an extra peak at 900-1000 K following ERDA and/or NRA measurements. The peak height appears to correlate with the amount of D initially trapped beyond the calculated IBA probe beam peak damage depth. Similar to pre-implantation damage scenarios, the IBA probe beam creates empty high energy traps which later retrap D atoms during TDS heating, which is supported by modelling experimental results using the Tritium Migration Analysis Program.

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.

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.

An Enriched Shell Finite Element for Progressive Damage Simulation in Composite Laminates

NASA Technical Reports Server (NTRS)

McElroy, Mark W.

2016-01-01

A formulation is presented for an enriched shell nite element capable of progressive damage simulation in composite laminates. The element uses a discrete adaptive splitting approach for damage representation that allows for a straightforward model creation procedure based on an initially low delity mesh. The enriched element is veri ed for Mode I, Mode II, and mixed Mode I/II delamination simulation using numerical benchmark data. Experimental validation is performed using test data from a delamination-migration experiment. Good correlation was found between the enriched shell element model results and the numerical and experimental data sets. The work presented in this paper is meant to serve as a rst milestone in the enriched element's development with an ultimate goal of simulating three-dimensional progressive damage processes in multidirectional laminates.

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.

Material distribution in light water reactor-type bundles tested under severe accident conditions

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

Noack, V.; Hagen, S.J.L.; Hofmann, P.

1997-02-01

Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO{sub 2} pellets, UO{sub 2} full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B{sub 4}C to represent boiling water reactor and VVER-1000 fuel elements) were subjected to temperature transients up to 2,300 K. A special method was applied to determine the axial mass distribution of bundle materials. The low-temperature melt formation by various interactions between zirconium and components of absorbermore » and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.« less

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.

Thermal Analysis of ZPPR High Pu Content Stored Fuel

DOE PAGES

Solbrig, Charles W.; Pope, Chad L.; Andrus, Jason P.

2014-09-17

The Zero Power Physics Reactor (ZPPR) operated from April 18, 1969, until 1990. ZPPR operated at low power for testing nuclear reactor designs. This paper examines the temperature of Pu content ZPPR fuel while it is in storage. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible cladding damage. Damage to the cladding could lead to fuel hydriding and oxidizing. A series of computer simulations were made to determine the range of temperatures potentially occuring in the ZPPR fuel. The maximum calculated fuel temperature is 292°C (558°F). Conservative assumptions in themore » model intentionally overestimate temperatures. The stored fuel temperatures are dependent on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel. Direct fuel temperatures could not be measured but storage bin doors, storage sleeve doors, and storage canister temperatures were measured. Comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are, as expected, higher than the actual temperatures. The maximum calculated fuel temperature with the most conservative assumptions is significantly below the fuel failure criterion of 600°C (1,112°F).« less

Molecular Dynamics Simulation of Fission Fragment Damage in Nuclear Fuel and Surrogate Material

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

Devanathan, Ram

ABSTRACT We have performed classical molecular dynamics simulations of swift heavy ion damage, typical of fission fragments, in nuclear fuel (UO 2) for energy deposition per unit length of 3.9 keV/nm. We did not observe amorphization. The damage mainly consisted of isolated point defects. Only about 1% of the displacements occur on the uranium sublattice. Oxygen Frenkel pairs are an order of magnitude more numerous than uranium Frenkel pairs in the primary damage state. In contrast, previous results show that the ratio of Frenkel pairs on the two sublattices is close to the stoichiometric ratio in ceria. These differences inmore » the primary damage state may lead to differences in radiation response of UO 2and CeO 2.« less

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.

Prescribed burning in ponderosa pine: fuel reductions and redistributing fuels near boles to prevent injury

Treesearch

Robert A. Progar; Kathryn H. Hrinkevich; Edward S. Clark; Matthew J. Rinella

2017-01-01

Fire suppression and other factors have resulted in high wildfire risk in the western US, and prescribed burning can be an effective tool for thinning forests and reducing fuels to lessen wildfire risks. However, prescribed burning sometimes fails to substantially reduce fuels and sometimes damages and kills valuable, large trees. This study compared fuel reductions...

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)

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

Detection of Earthquake-Induced Damage in a Framed Structure Using a Finite Element Model Updating Procedure

PubMed Central

Kim, Seung-Nam; Park, Taewon; Lee, Sang-Hyun

2014-01-01

Damage of a 5-story framed structure was identified from two types of measured data, which are frequency response functions (FRF) and natural frequencies, using a finite element (FE) model updating procedure. In this study, a procedure to determine the appropriate weightings for different groups of observations was proposed. In addition, a modified frame element which included rotational springs was used to construct the FE model for updating to represent concentrated damage at the member ends (a formulation for plastic hinges in framed structures subjected to strong earthquakes). The results of the model updating and subsequent damage detection when the rotational springs (RS model) were used were compared with those obtained using the conventional frame elements (FS model). Comparisons indicated that the RS model gave more accurate results than the FS model. That is, the errors in the natural frequencies of the updated models were smaller, and the identified damage showed clearer distinctions between damaged and undamaged members and was more consistent with observed damage. PMID:24574888

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

14 CFR 25.957 - Flow between interconnected tanks.

Code of Federal Regulations, 2013 CFR

2013-01-01

... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.957 Flow between interconnected tanks. If fuel can be pumped from one tank to another in flight, the fuel tank vents and the fuel transfer system must be designed so that no structural damage to the tanks can occur because of overfilling. ...

14 CFR 25.957 - Flow between interconnected tanks.

Code of Federal Regulations, 2011 CFR

2011-01-01

... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.957 Flow between interconnected tanks. If fuel can be pumped from one tank to another in flight, the fuel tank vents and the fuel transfer system must be designed so that no structural damage to the tanks can occur because of overfilling. ...

14 CFR 25.957 - Flow between interconnected tanks.

Code of Federal Regulations, 2010 CFR

2010-01-01

... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.957 Flow between interconnected tanks. If fuel can be pumped from one tank to another in flight, the fuel tank vents and the fuel transfer system must be designed so that no structural damage to the tanks can occur because of overfilling. ...

14 CFR 25.957 - Flow between interconnected tanks.

Code of Federal Regulations, 2014 CFR

2014-01-01

... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.957 Flow between interconnected tanks. If fuel can be pumped from one tank to another in flight, the fuel tank vents and the fuel transfer system must be designed so that no structural damage to the tanks can occur because of overfilling. ...

14 CFR 25.957 - Flow between interconnected tanks.

Code of Federal Regulations, 2012 CFR

2012-01-01

... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.957 Flow between interconnected tanks. If fuel can be pumped from one tank to another in flight, the fuel tank vents and the fuel transfer system must be designed so that no structural damage to the tanks can occur because of overfilling. ...

Early intervention reduces morbidity in extravasation injuries from 'lighter fuel' injection.

PubMed

Thaha, M A; McKinnell, T H; Graham, K E; Naasan, A N

2007-01-01

Injection of 'lighter fuel' with suicidal intent is rare. Extravasation of the chemical may rarely cause systemic toxicity, but usually it results in extensive soft tissue damage. Such injuries when managed by the traditional expectant policy are associated with considerable morbidity. Early aggressive surgical management using 'saline flush out' limits the tissue damage by stopping the natural progression of the chemical mediated injury and the subsequent inflammatory response, thereby allowing better skin preservation and functional outcome in these cases. We report a case of 'lighter fuel' subcutaneous extravasation injury managed by 'saline flush out' technique soon after presentation.

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

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.

Numerical-experimental investigation of load paths in DP800 dual phase steel during Nakajima test

NASA Astrophysics Data System (ADS)

Bergs, Thomas; Nick, Matthias; Feuerhack, Andreas; Trauth, Daniel; Klocke, Fritz

2018-05-01

Fuel efficiency requirements demand lightweight construction of vehicle body parts. The usage of advanced high strength steels permits a reduction of sheet thickness while still maintaining the overall strength required for crash safety. However, damage, internal defects (voids, inclusions, micro fractures), microstructural defects (varying grain size distribution, precipitates on grain boundaries, anisotropy) and surface defects (micro fractures, grooves) act as a concentration point for stress and consequently as an initiation point for failure both during deep drawing and in service. Considering damage evolution in the design of car body deep drawing processes allows for a further reduction in material usage and therefore body weight. Preliminary research has shown that a modification of load paths in forming processes can help mitigate the effects of damage on the material. This paper investigates the load paths in Nakajima tests of a DP800 dual phase steel to research damage in deep drawing processes. Investigation is done via a finite element model using experimentally validated material data for a DP800 dual phase steel. Numerical simulation allows for the investigation of load paths with respect to stress states, strain rates and temperature evolution, which cannot be easily observed in physical experiments. Stress triaxiality and the Lode parameter are used to describe the stress states. Their evolution during the Nakajima tests serves as an indicator for damage evolution. The large variety of sheet metal forming specific load paths in Nakajima tests allows a comprehensive evaluation of damage for deep drawing. The results of the numerical simulation conducted in this project and further physical experiments will later be used to calibrate a damage model for simulation of deep drawing processes.

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

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.

Investigations of Multiple Swirl-Venturi Fuel Injector Concepts: Recent Experimental Optical Measurement Results for 1-Point, 7-Point, and 9-Point Configurations

NASA Technical Reports Server (NTRS)

Hicks, Yolanda R.; Anderson, Robert C.; Tedder, Sarah A.; Tacina, Kathleen M.

2015-01-01

This paper presents results obtained during testing in optically-accessible, JP8-fueled, flame tube combustors using swirl-venturi lean direct injection (LDI) research hardware. The baseline LDI geometry has 9 fuel/air mixers arranged in a 3 x 3 array within a square chamber. 2-D results from this 9-element array are compared to results obtained in a cylindrical combustor using a 7-element array and a single element. In each case, the baseline element size remains the same. The effect of air swirler angle, and element arrangement on the presence of a central recirculation zone are presented. Only the highest swirl number air swirler produced a central recirculation zone for the single element swirl-venturi LDI and the 9-element LDI, but that same swirler did not produce a central recirculation zone for the 7-element LDI, possibly because of strong interactions due to element spacing within the array.

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.

View of damaged Apollo 13 Service Module from the Lunar/Command Modules

NASA Image and Video Library

1970-04-17

This view of the damaged Apollo 13 Service Module (SM) was photographed from the Lunar Module/Command Module following SM jettisoning. As seen here, an entire panel on the SM was blown away by the apparent explosion of oxygen tank number two located in Sector 4 of the SM. Two of the three fuel cells are visible just forward (above) the heavily damaged area. Three fuel cells, two oxygen tanks, and two hydrogen tanks are locate in Sector 4. The damaged area is located above the S-band high gain antenna. Nearest the camera is the Service Propulsion System (SPS) engine and nozzle. The damage to the SM caused the Apollo 13 crewmen to use the Lunar Module (LM) as a "lifeboat". The LM was jettisoned just prior to Earth reentry by the Command Module.

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.

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].

Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

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. Prototypical fuel elements 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 in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

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.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

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.

Enhancement of deuterium retention in damaged tungsten by plasma-induced defect clustering

NASA Astrophysics Data System (ADS)

Jin, Younggil; Roh, Ki-Baek; Sheen, Mi-Hyang; Kim, Nam-Kyun; Song, Jaemin; Kim, Young-Woon; Kim, Gon-Ho

2017-12-01

The enhancement of deuterium retention was investigated for tungsten in the presence of both 2.8 MeV self-ion induced cascade damage and fuel hydrogen isotope plasma. Vacancy clustering in cascade damaged polycrystalline tungsten occurred due to deuterium irradiation and was observed near the grain boundary by using all-step transmission electron microscopy analysis. Analysis of the highest desorption temperature peak using thermal desorption spectroscopy supports reasonable evidence of defect clustering in the damaged polycrystalline tungsten. The defect clustering was neither observed on the damaged polycrystalline tungsten without deuterium irradiation nor on the damaged single-crystalline tungsten with deuterium irradiation. This result implies the synergetic role of deuterium and grain boundary on defect clustering. This study proposes a path for the defect transform from point defect to defect cluster, by the agglomeration between irradiated deuterium and cascade damage-induced defect. This agglomeration may induce more severe damage on the tungsten divertor at which the high fuel hydrogen ions, fast neutrons, and self-ions are irradiated simultaneously and it would increase the in-vessel tritium inventory.

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)

Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

NASA Astrophysics Data System (ADS)

Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

2014-08-01

Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

TMI-2 (Three Mile Island Unit 2) core region defueling

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

Rodabaugh, J.M.; Cowser, D.K.

1988-01-01

In July of 1982, a video camera was inserted into the Three Mile Island Unit 2 reactor vessel providing the first visual evidence of core damage. This inspection, and numerous subsequent data acquisition tasks, revealed a central void /approx/1.5 m (5 ft) deep. This void region was surrounded by partial length fuel assemblies and ringed on the periphery by /approx/40 full-length, but partial cross-section, fuel assemblies. All of the original 177 fuel assemblies exhibited signs of damage. The bottom of the void cavity was covered with a bed of granular rubble, fuel assembly upper end fittings, control rod spiders, fuelmore » rod fragments, and fuel pellets. It was obvious that the normal plant refueling system not suitable for removing the damaged core. A new system of defueling tools and equipment was necessary to perform this task. Design of the new system was started immediately, followed by >1 yr of fabrication. Delivery and checkout of the defueling system occurred in mid-1985. Actual defueling was initiated in late 1985 with removal of the debris bed at the bottom of the core void. Obstructions to the debris, such as end fittings and fuel rod fragments ere removed first; then /approx/23,000 kg (50,000lb) of granular debris was quickly loaded into canisters. Core region defueling was completed in late 1987, /approx/2 yr after it was initiated.« less

KSC-2010-4791

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers at NASA's Michoud Assembly Facility in New Orleans prepare the Space Shuttle Program's last external fuel tank, ET-122, for transportation to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea secured aboard the Pegasus Barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4802

NASA Image and Video Library

2010-09-21

NEW ORLEANS -- At NASA's Michoud Assembly Facility in New Orleans the Space Shuttle Program's last external fuel tank, ET-122, is ready for transportation to NASA's Kennedy Space Center in Florida. Secured aboard the Pegasus Barge the tank will travel 900 miles by sea before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4892

NASA Image and Video Library

2010-09-28

CAPE CANAVERAL, Fla. -- The Space Shuttle Program's last external fuel tank, ET-122, moves from the Turn Basin to the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Once inside the Vehicle Assembly Building, it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the shuttle program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4812

NASA Image and Video Library

2010-09-22

LOUISIANA -- In Gulfport, La., workers connect the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, to Freedom Star, NASA's solid rocket booster retrieval ship. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4891

NASA Image and Video Library

2010-09-28

CAPE CANAVERAL, Fla. -- The Space Shuttle Program's last external fuel tank, ET-122, moves from the Turn Basin to the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Once inside the Vehicle Assembly Building, it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the shuttle program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4806

NASA Image and Video Library

2010-09-21

NEW ORLEANS -- A tug boat is pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, from NASA's Michoud Assembly Facility in New Orleans to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4895

NASA Image and Video Library

2010-09-28

CAPE CANAVERAL, Fla. -- The Space Shuttle Program's last external fuel tank, ET-122, enters the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans to Kennedy's Turn Basin aboard the Pegasus Barge. The tank eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the shuttle program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4797

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers escort the Space Shuttle Program's last external fuel tank, ET-122, from NASA's Michoud Assembly Facility in New Orleans onto the Pegasus Barge. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida secured aboard the barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4890

NASA Image and Video Library

2010-09-28

CAPE CANAVERAL, Fla. -- The Space Shuttle Program's last external fuel tank, ET-122, moves from the Turn Basin to the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Once inside the Vehicle Assembly Building, it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the shuttle program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4804

NASA Image and Video Library

2010-09-21

NEW ORLEANS -- A tug boat pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, from NASA's Michoud Assembly Facility in New Orleans to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4792

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers escort the Space Shuttle Program's last external fuel tank, ET-122, from NASA's Michoud Assembly Facility in New Orleans for transportation to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea secured aboard the Pegasus Barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4897

NASA Image and Video Library

2010-09-28

CAPE CANAVERAL, Fla. -- The Space Shuttle Program's last external fuel tank, ET-122, has been moved inside the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans to Kennedy's Turn Basin aboard the Pegasus Barge. The tank eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the shuttle program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4896

NASA Image and Video Library

2010-09-28

CAPE CANAVERAL, Fla. -- The Space Shuttle Program's last external fuel tank, ET-122, moves into the Vehicle Assembly Building at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans to Kennedy's Turn Basin aboard the Pegasus Barge. The tank eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the shuttle program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

Characterizing the Conductivity and Enhancing the Piezoresistivity of Carbon Nanotube-Polymeric Thin Films

PubMed Central

Zhao, Yingjun; Schagerl, Martin; Viechtbauer, Christoph

2017-01-01

The concept of lightweight design is widely employed for designing and constructing aerospace structures that can sustain extreme loads while also being fuel-efficient. Popular lightweight materials such as aluminum alloy and fiber-reinforced polymers (FRPs) possess outstanding mechanical properties, but their structural integrity requires constant assessment to ensure structural safety. Next-generation structural health monitoring systems for aerospace structures should be lightweight and integrated with the structure itself. In this study, a multi-walled carbon nanotube (MWCNT)-based polymer paint was developed to detect distributed damage in lightweight structures. The thin film’s electromechanical properties were characterized via cyclic loading tests. Moreover, the thin film’s bulk conductivity was characterized by finite element modeling. PMID:28773084

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.

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

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)

A Computational Efficient Physics Based Methodology for Modeling Ceramic Matrix Composites (Preprint)

DTIC Science & Technology

2011-11-01

elastic range, and with some simple forms of progressing damage . However, a general physics-based methodology to assess the initial and lifetime... damage evolution in the RVE for all possible load histories. Microstructural data on initial configuration and damage progression in CMCs were...the damaged elements will have changed, hence, a progressive damage model. The crack opening for each crack type in each element is stored as a

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.

Apollo 13 Service Module

NASA Image and Video Library

1970-04-17

AS13-59-8500A (17 April 1970) --- This view of the severely damaged Apollo 13 Service Module (SM) was photographed from the Lunar Module/Command Module (LM/CM) following SM jettisoning. As seen in this cropped image, enlarged to provide a close-up view of the damaged area, an entire panel on the SM was blown away by the apparent explosion of oxygen tank number two located in Sector 4 of the SM. Two of the three fuel cells are visible just forward (above) the heavily damaged area. Three fuel cells, two oxygen tanks, and two hydrogen tanks are located in Sector 4. The damaged area is located above the S-Band high gain antenna. Nearest the camera is the Service Propulsion System (SPS) engine and nozzle. The damage to the SM caused the Apollo 13 crew members to use the LM as a "lifeboat". The LM was jettisoned just prior to Earth re-entry by the CM. Photo credit: NASA

View of damaged Apollo 13 Service Module from the Lunar/Command Modules

NASA Image and Video Library

1970-04-17

AS13-59-8501 (17 April 1970) --- This view of the severely damaged Apollo 13 Service Module (SM) was photographed from the Lunar Module/Command Module (LM/CM) following SM jettisoning. As seen here, an entire panel on the SM was blown away by the apparent explosion of oxygen tank number two located in Sector 4 of the SM. Two of the three fuel cells are visible just forward (above) the heavily damaged area. Three fuel cells, two oxygen tanks, and two hydrogen tanks are located in Sector 4. The damaged area is located above the S-Band high gain antenna. Nearest the camera is the Service Propulsion System (SPS) engine and nozzle. The damage to the SM caused the Apollo 13 crew men to use the LM as a "lifeboat." The LM was jettisoned just prior to Earth re-entry by the CM.

Surface fuel changes after severe disturbances in northern Rocky Mountain ecosystems

Treesearch

Chris Stalling; Robert E. Keane; Molly Retzlaff

2017-01-01

It is generally assumed that severe disturbances predispose damaged forests to high fire hazard by creating heavy fuel loading conditions. Of special concern is the perception that surface fuel loadings become high as recently killed trees deposit foliage and woody material on the ground and that these high fuel loadings may cause abnormally severe fires. This study...

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

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.

A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144Ce as a Surrogate for Pu Accountancy

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

D. L. Chichester; S. J. Thompson

2013-09-01

This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium inmore » the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for using cerium, which is rather easy to analyze using passive nondestructive analysis gamma-ray spectrometry, as a surrogate for plutonium in the final analysis of TMI-2 melted fuel debris. The generation of this report is motivated by the need to perform nuclear material accountancy measurements on the melted fuel debris that will be excavated from the damaged nuclear reactors at the Fukushima Daiichi nuclear power plant in Japan, which were destroyed by the Tohoku earthquake and tsunami on March 11, 2011. Lessons may be taken from prior U.S. work related to the study of the TMI-2 core debris to support the development of new assay methods for use at Fukushima Daiichi. While significant differences exist between the two reactor systems (pressurized water reactor (TMI-2) versus boiling water reactor (FD), fresh water post-accident cooing (TMI-2) versus salt water (FD), maintained containment (TMI-2) versus loss of containment (FD)) there remain sufficient similarities to motivate these comparisons.« less

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

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.

Simulation of laminate composites degradation using mesoscopic non-local damage model and non-local layered shell element

NASA Astrophysics Data System (ADS)

Germain, Norbert; Besson, Jacques; Feyel, Frédéric

2007-07-01

Simulating damage and failure of laminate composites structures often fails when using the standard finite element procedure. The difficulties arise from an uncontrolled mesh dependence caused by damage localization and an increase in computational costs. One of the solutions to the first problem, widely used to predict the failure of metallic materials, consists of using non-local damage constitutive equations. The second difficulty can then be solved using specific finite element formulations, such as shell element, which decrease the number of degrees of freedom. The main contribution of this paper consists of extending these techniques to layered materials such as polymer matrix composites. An extension of the non-local implicit gradient formulation, accounting for anisotropy and stratification, and an original layered shell element, based on a new partition of the unity, are proposed. Finally the efficiency of the resulting numerical scheme is studied by comparing simulation with experimental results.

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.

A damage mechanics based general purpose interface/contact element

NASA Astrophysics Data System (ADS)

Yan, Chengyong

Most of the microelectronics packaging structures consist of layered substrates connected with bonding materials, such as solder or epoxy. Predicting the thermomechanical behavior of these multilayered structures is a challenging task in electronic packaging engineering. In a layered structure the most complex part is always the interfaces between the strates. Simulating the thermo-mechanical behavior of such interfaces, is the main theme of this dissertation. The most commonly used solder material, Pb-Sn alloy, has a very low melting temperature 180sp°C, so that the material demonstrates a highly viscous behavior. And, creep usually dominates the failure mechanism. Hence, the theory of viscoplasticity is adapted to describe the constitutive behavior. In a multilayered assembly each layer has a different coefficient of thermal expansion. Under thermal cycling, due to heat dissipated from circuits, interfaces and interconnects experience low cycle fatigue. Presently, the state-of-the art damage mechanics model used for fatigue life predictions is based on Kachanov (1986) continuum damage model. This model uses plastic strain as a damage criterion. Since plastic strain is a stress path dependent value, the criterion does not yield unique damage values for the same state of stress. In this dissertation a new damage evolution equation based on the second law of thermodynamic is proposed. The new criterion is based on the entropy of the system and it yields unique damage values for all stress paths to the final state of stress. In the electronics industry, there is a strong desire to develop fatigue free interconnections. The proposed interface/contact element can also simulate the behavior of the fatigue free Z-direction thin film interconnections as well as traditional layered interconnects. The proposed interface element can simulate behavior of a bonded interface or unbonded sliding interface, also called contact element. The proposed element was verified against laboratory test data presented in the literature. The results demonstrate that the proposed element and the damage law perform very well. The most important scientific contribution of this dissertation is the proposed damage criterion based on second law of thermodynamic and entropy of the system. The proposed general purpose interface/contact element is another contribution of this research. Compared to the previous adhoc interface elements proposed in the literature, the new one is, much more powerful and includes creep, plastic deformations, sliding, temperature, damage, cyclic behavior and fatigue life in a unified formulation.

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.

Rotor damage detection by using piezoelectric impedance

NASA Astrophysics Data System (ADS)

Qin, Y.; Tao, Y.; Mao, Y. F.

2016-04-01

Rotor is a core component of rotary machinery. Once the rotor has the damage, it may lead to a major accident. Thus the quantitative rotor damage detection method based on piezoelectric impedance is studied in this paper. With the governing equation of piezoelectric transducer (PZT) in a cylindrical coordinate, the displacement along the radius direction is derived. The charge of PZT is calculated by the electric displacement. Then, by the use of the obtained displacement and charge, an analytic piezoelectric impedance model of the rotor is built. Given the circular boundary condition of a rotor, annular elements are used as the analyzed objects and spectral element method is used to set up the damage detection model. The Electro-Mechanical (E/M) coupled impedance expression of an undamaged rotor is deduced with the application of a low-cost impedance test circuit. A Taylor expansion method is used to obtain the approximate E/M coupled impedance expression for the damaged rotor. After obtaining the difference between the undamaged and damaged rotor impedance, a rotor damage detection method is proposed. This method can directly calculate the change of bending stiffness of the structural elements, it follows that the rotor damage can be effectively detected. Finally, a preset damage configuration is used for the numerical simulation. The result shows that the quantitative damage detection algorithm based on spectral element method and piezoelectric impedance proposed in this paper can identify the location and the severity of the damaged rotor accurately.

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.

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.

Electron beam induced radiation damage in the catalyst layer of a proton exchange membrane fuel cell.

PubMed

He, Qianping; Chen, Jihua; Keffer, David J; Joy, David C

2014-01-01

Electron microscopy is an essential tool for the evaluation of microstructure and properties of the catalyst layer (CL) of proton exchange membrane fuel cells (PEMFCs). However, electron microscopy has one unavoidable drawback, which is radiation damage. Samples suffer temporary or permanent change of the surface or bulk structure under radiation damage, which can cause ambiguity in the characterization of the sample. To better understand the mechanism of radiation damage of CL samples and to be able to separate the morphological features intrinsic to the material from the consequences of electron radiation damage, a series of experiments based on high-angle annular dark-field-scanning transmission scanning microscope (HAADF-STEM), energy filtering transmission scanning microscope (EFTEM), and electron energy loss spectrum (EELS) are conducted. It is observed that for thin samples (0.3-1 times λ), increasing the incident beam energy can mitigate the radiation damage. Platinum nanoparticles in the CL sample facilitate the radiation damage. The radiation damage of the catalyst sample starts from the interface of Pt/C or defective thin edge and primarily occurs in the form of mass loss accompanied by atomic displacement and edge curl. These results provide important insights on the mechanism of CL radiation damage. Possible strategies of mitigating the radiation damage are provided. © 2013 Wiley Periodicals, Inc.

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

Residual life and strength estimates of aircraft structural components with MSD/MED

NASA Technical Reports Server (NTRS)

Singh, Ripudaman; Park, Jai H.; Atluri, Satya N.

1994-01-01

Economic and safe operation of the flight vehicles flying beyond their initial design life calls for an in-depth structural integrity evaluation of all components with potential for catastrophic damages. Fuselage panels with cracked skin and/or stiffening elements is one such example. A three level analytical approach is developed to analyze the pressurized fuselage stiffened shell panels with damaged skin or stiffening elements. A global finite element analysis is first carried out to obtain the load flow pattern through the damaged panel. As an intermediate step, the damaged zone is treated as a spatially three-dimensional structure modeled by plate and shell finite elements, with all the neighboring elements that can alter the stress state at the crack tip. This is followed by the Schwartz-Neumann alternating method for local analysis to obtain the relevant crack tip parameters that govern the onset of fracture and the crack growth. The methodology developed is generic in nature and aims at handling a large fraction of problem areas identified by the Industry Committee on Wide-Spread Fatigue Damage.

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.

Simulating fuel reduction scenarios on a wildland-urban interface in northeastern Oregon.

Treesearch

Alan A. Ager; R. James Barbour; Jane L. Hayes

2005-01-01

We analyzed the long-term effects of fuels reduction treatments around a wildland-urban interface located in the Blue Mountains near La Grande, Oregon. The study area is targeted for fuels reduction treatments on both private and federal lands to reduce the risk of severe wildfire and associated damage to property and homes. We modeled a number of hypothetical fuel...

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.

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

Conversion of a Micro, Glow-Ignition, Two-Stroke Engine from Nitromethane-Methanol Blend Fuel to Military Jet Propellant (JP-8)

DTIC Science & Technology

2012-01-01

Table 10-4: Selected Birk polyimide heater sizes, resistances and locations [37] ........................ 79 Table 10-5: Final starting tests with (3...damage, and fire are prevalent. Kerosene type fuels are also cheaper and more common than nitromethane-methanol blend fuels. One final note is...diesel fuel was changed to produce lower emissions, the abrasiveness of diesel fuel increased. This was especially problematic for the new high

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.

43 CFR 5003.1 - Effect of decisions; general.

Code of Federal Regulations, 2013 CFR

2013-10-01

... risk of wildfire due to drought, fuels buildup, or other reasons, or at immediate risk of erosion or other damage due to wildfire, BLM may make a wildfire management decision made under this part and parts.... Wildfire management includes but is not limited to: (1) Fuel reduction or fuel treatment such as prescribed...

43 CFR 5003.1 - Effect of decisions; general.

Code of Federal Regulations, 2012 CFR

2012-10-01

... risk of wildfire due to drought, fuels buildup, or other reasons, or at immediate risk of erosion or other damage due to wildfire, BLM may make a wildfire management decision made under this part and parts.... Wildfire management includes but is not limited to: (1) Fuel reduction or fuel treatment such as prescribed...

43 CFR 5003.1 - Effect of decisions; general.

Code of Federal Regulations, 2014 CFR

2014-10-01

... risk of wildfire due to drought, fuels buildup, or other reasons, or at immediate risk of erosion or other damage due to wildfire, BLM may make a wildfire management decision made under this part and parts.... Wildfire management includes but is not limited to: (1) Fuel reduction or fuel treatment such as prescribed...

43 CFR 5003.1 - Effect of decisions; general.

Code of Federal Regulations, 2011 CFR

2011-10-01

... risk of wildfire due to drought, fuels buildup, or other reasons, or at immediate risk of erosion or other damage due to wildfire, BLM may make a wildfire management decision made under this part and parts.... Wildfire management includes but is not limited to: (1) Fuel reduction or fuel treatment such as prescribed...

40 CFR 69.52 - Non-motor vehicle diesel fuel.

Code of Federal Regulations, 2014 CFR

2014-07-01

... diesel vehicles and engines Its use may damage these vehicles and engines. For use in all other diesel vehicles and engines. (ii) 15 ppm sulfur diesel fuel. From June 1, 2006 through May 31, 2010. ULTRA-LOW... and engines. Recommended for use in all diesel vehicles and engines. (iii) 15 ppm sulfur diesel fuel...

Physics From the News -- Fukushima Daiichi: Radiation Doses and Dose Rates

NASA Astrophysics Data System (ADS)

Bartlett, A. A.

2011-09-01

The nuclear disaster that was triggered by the Japanese earthquake and the following tsunami of March 11, 2011, continues to be the subject of a great deal of news coverage. The tsunami caused severe damage to the nuclear power reactors at Fukushima Daiichi, and this led to the escape of unknown quantities of radioactive material from the damaged fuel rods in the reactors and from the associated storage facilities for the fuel rods that had been removed from the reactors.

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.

Saturn Apollo Program

NASA Image and Video Library

1970-04-01

Apollo 13 onboard photo: This view of the severely damaged Apollo 13 Service Module was photographed from the Lunar Module/Command Module following the jettison of the Service Module. As seen here, an entire panel of the Service Module was blown away by the apparent explosion of oxygen tank number two located in Sector 4 of the Service Module. Two of the three fuel cells are visible just forward (above) the heavily damaged area. Three fuel cells, two oxygen tanks, and two hydrogen tanks, are located in Sector 4. The damaged area is located above the S-band high gain anterna. Nearest the camera is the Service Propulsion System (SPS) engine and nozzle. The damage to the Service Module caused the Apollo 13 crewmen to use the Lunar Module as a lifeboat. The Lunar Module was jettisoned by the Command Module just prior to Earth re-entry.

ANALYSIS OF BORON DILUTION TRANSIENTS IN PWRS.

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

DIAMOND,D.J.BROMLEY,B.P.ARONSON,A.L.

2004-02-04

A study has been carried out with PARCS/RELAP5 to understand the consequences of hypothetical boron dilution events in pressurized water reactors. The scenarios of concern start with a small-break loss-of-coolant accident. If the event leads to boiling in the core and then the loss of natural circulation, a boron-free condensate can accumulate in the cold leg. The dilution event happens when natural circulation is re-established or a reactor coolant pump (RCP) is restarted in violation of operating procedures. This event is of particular concern in B&W reactors with a lowered-loop design and is a Generic Safety Issue for the U.S.more » Nuclear Regulatory Commission. The results of calculations with the reestablishment of natural circulation show that there is no unacceptable fuel damage. This is determined by calculating the maximum fuel pellet enthalpy, based on the three-dimensional model, and comparing it with the criterion for damage. The calculation is based on a model of a B&W reactor at beginning of the fuel cycle. If an RCP is restarted, unacceptable fuel damage may be possible in plants with sufficiently large volumes of boron-free condensate in the cold leg.« less

14 CFR 23.1337 - Powerplant instruments installation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... against damage; (3) Each sight gauge that forms a trap in which water can collect and freeze must have... not proceeding as planned. (c) Fuel flowmeter system. If a fuel flowmeter system is installed, each...

14 CFR 23.1337 - Powerplant instruments installation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... against damage; (3) Each sight gauge that forms a trap in which water can collect and freeze must have... not proceeding as planned. (c) Fuel flowmeter system. If a fuel flowmeter system is installed, each...

14 CFR 23.1337 - Powerplant instruments installation.

Code of Federal Regulations, 2010 CFR

2010-01-01

... against damage; (3) Each sight gauge that forms a trap in which water can collect and freeze must have... not proceeding as planned. (c) Fuel flowmeter system. If a fuel flowmeter system is installed, each...

14 CFR 23.1337 - Powerplant instruments installation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... against damage; (3) Each sight gauge that forms a trap in which water can collect and freeze must have... not proceeding as planned. (c) Fuel flowmeter system. If a fuel flowmeter system is installed, each...

14 CFR 23.1337 - Powerplant instruments installation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... against damage; (3) Each sight gauge that forms a trap in which water can collect and freeze must have... not proceeding as planned. (c) Fuel flowmeter system. If a fuel flowmeter system is installed, each...

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.

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.

KSC-2010-4852

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- The Pegasus Barge, carrying the Space Shuttle Program's last external fuel tank, ET-122, nears NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett

KSC-2010-4865

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tugboat pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jim Grossmann

KSC-2010-4839

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, to the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4840

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, to the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4876

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- The Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, arrives at the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. Next, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4813

NASA Image and Video Library

2010-09-22

GULFPORT, La. -- At Gulfport, La., Michael Nicholas, captain M/V Freedom Star, guides NASA's solid rocket booster retrieval ship out of port pulling the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4862

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- NASA's Pegasus barge, carrying the Space Shuttle Program's last external fuel tank, ET-122, arrives at the Turn Basin of NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. Next, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett

KSC-2010-4819

NASA Image and Video Library

2010-09-25

CAPE CANAVERAL, Fla. -- This sunrise view from the stern of Freedom Star, one of NASA's solid rocket booster retrieval ships, shows the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4793

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- To commemorate the history of the Space Shuttle Program's last external fuel tank, its intertank door is emblazoned with an ET-122 insignia. The external tank will travel 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans to NASA's Kennedy Space Center in Florida secured aboard the Pegasus Barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4836

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, toward the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb., 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4826

NASA Image and Video Library

2010-09-26

CAPE CANAVERAL, Fla. -- Deckhands on Freedom Star, one of NASA's solid rocket booster retrieval ships, keep the ship in good repair as it pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4874

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- The Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, arrives at the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. Next, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4799

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers watch the progress of the Space Shuttle Program's last external fuel tank, ET-122, at NASA's Michoud Assembly Facility in New Orleans, as it is being loaded onto the Pegasus Barge. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida secured aboard the barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4841

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, to the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4833

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, toward the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4803

NASA Image and Video Library

2010-09-21

NEW ORLEANS -- At NASA's Michoud Assembly Facility in New Orleans a tug boat is prepared to escort the Space Shuttle Program's last external fuel tank, ET-122, for transportation to NASA's Kennedy Space Center in Florida. Secured aboard the Pegasus Barge the tank will travel 900 miles by sea before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4801

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers check the progress of the Space Shuttle Program's last external fuel tank, ET-122, at NASA's Michoud Assembly Facility in New Orleans as it is being loaded onto the Pegasus Barge. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida secured aboard the barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4838

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- At NASA's Kennedy Space Center in Florida, the Pegasus Barge, carrying the Space Shuttle Program's last external fuel tank, ET-122, arrives at the Turn Basin. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4824

NASA Image and Video Library

2010-09-26

CAPE CANAVERAL, Fla. -- This view is from the deck of Freedom Star, one of NASA's solid rocket booster retrieval ships, as it pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4816

NASA Image and Video Library

2010-09-22

CAPE CANAVERAL, Fla. -- This view from Freedom Star, one NASA's solid rocket booster retrieval ships, shows the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, as it is transported to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4827

NASA Image and Video Library

2010-09-26

CAPE CANAVERAL, Fla. -- This view from the stern of Freedom Star, one of NASA's solid rocket booster retrieval ships, shows the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4823

NASA Image and Video Library

2010-09-26

CAPE CANAVERAL, Fla. -- Deckhands on Freedom Star, one of NASA's solid rocket booster retrieval ships, keep the ship in good repair as it pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4815

NASA Image and Video Library

2010-09-22

CAPE CANAVERAL, Fla. -- This view from the stern of Freedom Star, one of NASA's solid rocket booster retrieval ships, shows the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, as it is transported to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea, offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4871

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tugboat pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, toward the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. Next, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4908

NASA Image and Video Library

2010-09-28

CAPE CANAVERAL, Fla. -- This overhead view shows the Space Shuttle Program's last external fuel tank, ET-122, as it is being transported to the Vehicle Assembly Building (VAB) at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea, carried in the Pegasus Barge, from NASA's Michoud Assembly Facility in New Orleans. Once inside the VAB, it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station targeted to launch Feb. 2011. STS-134 currently is scheduled to be the last mission in the shuttle program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kevin O'Connell

KSC-2010-4837

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, toward the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb., 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4820

NASA Image and Video Library

2010-09-25

CAPE CANAVERAL, Fla. -- This view from the stern of Freedom Star, one of NASA's solid rocket booster retrieval ships, shows the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4822

NASA Image and Video Library

2010-09-26

CAPE CANAVERAL, Fla. -- A deckhand on Freedom Star, one of NASA's solid rocket booster retrieval ships, keeps the ship in good repair as it pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4834

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, toward the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb., 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4835

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- A tug boat pulls the Space Shuttle Program's last external fuel tank, ET-122, toward the Turn Basin at NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. Next, the tank will be offloaded and moved to Kennedy's Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4821

NASA Image and Video Library

2010-09-26

CAPE CANAVERAL, Fla. -- Deckhands on Freedom Star, one of NASA's solid rocket booster retrieval ships, keep the ship in good repair as it pulls the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4798

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers monitor the progress of the Space Shuttle Program's last external fuel tank, ET-122, from NASA's Michoud Assembly Facility in New Orleans as it is being loaded onto the Pegasus Barge. The tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida secured aboard the barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4800

NASA Image and Video Library

2010-09-20

NEW ORLEANS -- Workers monitor the progress of the Space Shuttle Program's last external fuel tank, ET-122, at NASA's Michoud Assembly Facility in New Orleans as it is being loaded onto the Pegasus BargeThe tank will travel 900 miles by sea to NASA's Kennedy Space Center in Florida secured aboard the barge, offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

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.

Mixed-Mode Decohesion Elements for Analyses of Progressive Delamination

NASA Technical Reports Server (NTRS)

Davila, Carlos G.; Camanho, Pedro P.; deMoura, Marcelo F.

2001-01-01

A new 8-node decohesion element with mixed mode capability is proposed and demonstrated. The element is used at the interface between solid finite elements to model the initiation and propagation of delamination. A single displacement-based damage parameter is used in a strain softening law to track the damage state of the interface. The method can be used in conjunction with conventional material degradation procedures to account for inplane and intra-laminar damage modes. The accuracy of the predictions is evaluated in single mode delamination tests, in the mixed-mode bending test, and in a structural configuration consisting of the debonding of a stiffener flange from its skin.

Regulation of the yeast RAD2 gene: DNA damage-dependent induction correlates with protein binding to regulatory sequences and their deletion influences survival.

PubMed

Siede, W; Friedberg, E C

1992-03-01

In the yeast Saccharomyces cerevisiae the RAD2 gene is absolutely required for damage-specific incision of DNA during nucleotide excision repair and is inducible by DNA-damaging agents. In the present study we correlated sensitivity to killing by DNA-damaging agents with the deletion of previously defined specific promoter elements. Deletion of the element DRE2 increased the UV sensitivity of cells in both the G1/early S and S/G2 phases of the cell cycle as well as in stationary phase. On the other hand, increased UV sensitivity associated with deletion of the sequence-related element DRE1 was restricted to cells irradiated in G1/S. Specific binding of protein(s) to the promoter elements DRE1 and DRE2 was observed under non-inducing conditions using gel retardation assays. Exposure of cells to DNA-damaging agents resulted in increased protein binding that was dependent on de novo protein synthesis.

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.

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.

Method for storing spent nuclear fuel in repositories

DOEpatents

Schweitzer, Donald G.; Sastre, Cesar; Winsche, Warren

1981-01-01

A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

Method for storing spent nuclear fuel in repositories

DOEpatents

Schweitzer, D.G.; Sastre, C.; Winsche, W.

A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

State of Fukushima nuclear fuel debris tracked by Cs137 in cooling water.

PubMed

Grambow, B; Mostafavi, M

2014-11-01

It is still difficult to assess the risk originating from the radioactivity inventory remaining in the damaged Fukushima nuclear reactors. Here we show that cooling water analyses provide a means to assess source terms for potential future releases. Until now already about 34% of the inventories of (137)Cs of three reactors has been released into water. We found that the release rate of (137)Cs has been constant for 2 years at about 1.8% of the inventory per year indicating ongoing dissolution of the fuel debris. Compared to laboratory studies on spent nuclear fuel behavior in water, (137)Cs release rates are on the higher end, caused by the strong radiation field and oxidant production by water radiolysis and by impacts of accessible grain boundaries. It is concluded that radionuclide analyses in cooling water allow tracking of the conditions of the damaged fuel and the associated risks.

Spent fuel cask handling at an operating nuclear power plant

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

Pal, A.C.

1988-01-01

The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices atmore » all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant.« less

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)

Development of Nanomaterials for Nuclear Energetics

NASA Astrophysics Data System (ADS)

Petrunin, V. F.

Structure and properties peculiarities of the nanocrystalline powders give the opportunity to design new and to develop a modernization of nuclear energy industry materials. It was shown experimentally, that addition of 5-10% uranium dioxide nanocrystalline powder to traditional coarse powder allows to decrease the sintering temperature or to increase the fuel tablets size of grain. Similar perspectives for the technology of neutron absorbing tablets of control-rod modernization are shown by nanopowder of dysprosium hafnate changing instead now using boron carbide. It is powders in nanocrystalline state get an opportunity to sinter them and to receive compact tablet with 8,2-8,4 g/cm2 density for automatic defence system of nuclear reactor. Resource of dysprosium hafnate ceramics can be 18-20 years instead 4-5 years for boron carbide. To step up the radiation-damage stability of fuel element jacket material was suggested to strengthen a heat-resistant ferrite-martensite steel by Y2O3 nanocrystalline powder addition. Nanopowder with size of particles 560 nm and crystallite size 9 nm was prepeared by chemical coprecipitation method. To make lighter the container for transport and provisional disposal of exposed fuel from nuclear reactor a new boron-aluminium alloy called as boral was developed. This composite armed with nanopowders of boron-containing materials and heavy metals oxides can replace succesburnt-up corrosion-resistant steels.

Landscape Preferences, Amenity, and Bushfire Risk in New South Wales, Australia.

PubMed

Gill, Nicholas; Dun, Olivia; Brennan-Horley, Chris; Eriksen, Christine

2015-09-01

This paper examines landscape preferences of residents in amenity-rich bushfire-prone landscapes in New South Wales, Australia. Insights are provided into vegetation preferences in areas where properties neighbor large areas of native vegetation, such as national parks, or exist within a matrix of cleared and vegetated private and public land. In such areas, managing fuel loads in the proximity of houses is likely to reduce the risk of house loss and damage. Preferences for vegetation appearance and structure were related to varying fuel loads, particularly the density of understorey vegetation and larger trees. The study adopted a qualitative visual research approach, which used ranking and photo-elicitation as part of a broader interview. A visual approach aids in focusing on outcomes of fuel management interventions, for example, by using the same photo scenes to firstly derive residents' perceptions of amenity and secondly, residents' perceptions of bushfire risk. The results are consistent with existing research on landscape preferences; residents tend to prefer relatively open woodland or forest landscapes with good visual and physical access but with elements that provoke their interest. Overall, residents' landscape preferences were found to be consistent with vegetation management that reduces bushfire risk to houses. The terms in which preferences were expressed provide scope for agency engagement with residents in order to facilitate management that meets amenity and hazard reduction goals on private land.

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.

75 FR 43101 - Airworthiness Directives; PIAGGIO AERO INDUSTRIES S.p.A Model PIAGGIO P-180 Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-23

...) Code 79: Engine Oil. Reason (e) The mandatory continuing airworthiness information (MCAI) states: A... fuel heater caused a fuel leakage in the engine nacelle; investigation revealed that the damage to the fuel heater was due to chafing with an oil cooling system hose. PIAGGIO AERO INDUSTIRES (PAI) issued...

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...

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.

Disposal options for polluted plants grown on heavy metal contaminated brownfield lands - A review.

PubMed

Kovacs, Helga; Szemmelveisz, Katalin

2017-01-01

Reducing or preventing damage caused by environmental pollution is a significant goal nowadays. Phytoextraction, as remediation technique is widely used, but during the process, the heavy metal content of the biomass grown on these sites special treatment and disposal techniques are required, for example liquid extraction, direct disposal, composting, and combustion. These processes are discussed in this review in economical and environmental aspects. The following main properties are analyzed: form and harmful element content of remains, utilization of the main and byproducts, affect to the environment during the treatment and disposal. The thermal treatment (combustion, gasification) of contaminated biomass provides a promising alternative disposal option, because the energy production affects the rate of return, and the harmful elements are riched in a small amount of solid remains depending on the ash content of the plant (1-2%). The biomass combustion technology is a wildely used energy production process in residential and industrial scale, but the ordinary biomass firing systems are not suited to burn this type of fuel without environmental risk. Copyright © 2016 Elsevier Ltd. All rights reserved.

Damage identification of a reinforced concrete frame by finite element model updating using damage parameterization

NASA Astrophysics Data System (ADS)

Fang, Sheng-En; Perera, Ricardo; De Roeck, Guido

2008-06-01

This paper develops a sensitivity-based updating method to identify the damage in a tested reinforced concrete (RC) frame modeled with a two-dimensional planar finite element (FE) by minimizing the discrepancies of modal frequencies and mode shapes. In order to reduce the number of unknown variables, a bidimensional damage (element) function is proposed, resulting in a considerable improvement of the optimization performance. For damage identification, a reference FE model of the undamaged frame divided into a few damage functions is firstly obtained and then a rough identification is carried out to detect possible damage locations, which are subsequently refined with new damage functions to accurately identify the damage. From a design point of view, it would be useful to evaluate, in a simplified way, the remaining bending stiffness of cracked beam sections or segments. Hence, an RC damage model based on a static mechanism is proposed to estimate the remnant stiffness of a cracked RC beam segment. The damage model is based on the assumption that the damage effect spreads over a region and the stiffness in the segment changes linearly. Furthermore, the stiffness reduction evaluated using this damage model is compared with the FE updating result. It is shown that the proposed bidimensional damage function is useful in producing a well-conditioned optimization problem and the aforementioned damage model can be used for an approximate stiffness estimation of a cracked beam segment.

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

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 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). Last year NTREES was successfully used to satisfy a testing milestone for the Nuclear Cryogenic Propulsion Stage (NCPS) project and met or exceeded all required objectives.

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.

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

Numerical analysis of static strength for different damages of hydraulic structures when changing stressed and strained state

NASA Astrophysics Data System (ADS)

Volosukhin, V. A.; Bandurin, M. A.; Vanzha, V. V.; Mikheev, A. V.; Volosukhin, Y. V.

2018-05-01

The results of finite element state simulation of stressed and strained changes under different damages of hydraulic structures are presented. As a result of the experiment, a solidstate model of bearing elements was built. Stressed and strained state of reinforced concrete bearing elements under different load combinations is considered. Intensive threshold of danger to form longitudinal cracks and defects in reinforced concrete elements is determined.

MCNP-model for the OAEP Thai Research Reactor

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

Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculationsmore » were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.« less

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.

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.

75 FR 12464 - Airworthiness Directives; McDonnell Douglas Corporation Model MD-11 and MD-11F Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-16

... would require a one-time inspection to detect damage of the wire assemblies of the tail tank fuel system, a wiring change, and corrective actions if necessary. This proposed AD results from fuel system... existing maintenance practices for fuel tank systems. As a result of those findings, we issued a regulation...

Fuel treatment effectiveness in forests of the upper Atlantic Coastal Plain—an evaluation at two spatial scales

Treesearch

Roger D. Ottmar; Susan J. Prichard

2012-01-01

Fuel treatment effectiveness in Southern forests has been demonstrated using fire behavior modeling and observations of reduced wildfire area and tree damage. However, assessments of treatment effectiveness may be improved with a more rigorous accounting of the fuel characteristics. We present two case studies to introduce a relatively new approach to characterizing...

Fiber optic distributed chemical sensor for the real time detection of hydrocarbon fuel leaks

NASA Astrophysics Data System (ADS)

Mendoza, Edgar; Kempen, C.; Esterkin, Yan; Sun, Sunjian

2015-09-01

With the increase worldwide demand for hydrocarbon fuels and the vast development of new fuel production and delivery infrastructure installations around the world, there is a growing need for reliable hydrocarbon fuel leak detection technologies to provide safety and reduce environmental risks. Hydrocarbon leaks (gas or liquid) pose an extreme danger and need to be detected very quickly to avoid potential disasters. Gas leaks have the greatest potential for causing damage due to the explosion risk from the dispersion of gas clouds. This paper describes progress towards the development of a fast response, high sensitivity, distributed fiber optic fuel leak detection (HySense™) system based on the use of an optical fiber that uses a hydrocarbon sensitive fluorescent coating to detect the presence of fuel leaks present in close proximity along the length of the sensor fiber. The HySense™ system operates in two modes, leak detection and leak localization, and will trigger an alarm within seconds of exposure contact. The fast and accurate response of the sensor provides reliable fluid leak detection for pipelines, storage tanks, airports, pumps, and valves to detect and minimize any potential catastrophic damage.

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.

The Use of the STAGS Finite Element Code in Stitched Structures Development

NASA Technical Reports Server (NTRS)

Jegley, Dawn C.; Lovejoy, Andrew E.

2014-01-01

In the last 30 years NASA has worked in collaboration with industry to develop enabling technologies needed to make aircraft more fuel-efficient and more affordable. The focus on the airframe has been to reduce weight, improve damage tolerance and better understand structural behavior under realistic flight and ground loading conditions. Stitched structure is a technology that can address the weight savings, cost reduction, and damage tolerance goals, but only if it is supported by accurate analytical techniques. Development of stitched technology began in the 1990's as a partnership between NASA and Boeing (McDonnell Douglas at the time) under the Advanced Composites Technology Program and has continued under various titles and programs and into the Environmentally Responsible Aviation Project today. These programs contained development efforts involving manufacturing development, design, detailed analysis, and testing. Each phase of development, from coupons to large aircraft components was supported by detailed analysis to prove that the behavior of these structures was well-understood and predictable. The Structural Analysis of General Shells (STAGS) computer code was a critical tool used in the development of many stitched structures. As a key developer of STAGS, Charles Rankin's contribution to the programs was quite significant. Key features of STAGS used in these analyses and discussed in this paper include its accurate nonlinear and post-buckling capabilities, its ability to predict damage growth, and the use of Lagrange constraints and follower forces.

EXPERIMENTAL STUDIES OF TRANSIENT EFFECTS IN FAST REACTOR FUELS. SERIES I. UO$sub 2$ IRRADIATIONS

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

Field, J.H.

1962-11-15

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO/sub 2/ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel. (auth)

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...

Strain-Based Damage Determination Using Finite Element Analysis for Structural Health Management

NASA Technical Reports Server (NTRS)

Hochhalter, Jacob D.; Krishnamurthy, Thiagaraja; Aguilo, Miguel A.

2016-01-01

A damage determination method is presented that relies on in-service strain sensor measurements. The method employs a gradient-based optimization procedure combined with the finite element method for solution to the forward problem. It is demonstrated that strains, measured at a limited number of sensors, can be used to accurately determine the location, size, and orientation of damage. Numerical examples are presented to demonstrate the general procedure. This work is motivated by the need to provide structural health management systems with a real-time damage characterization. The damage cases investigated herein are characteristic of point-source damage, which can attain critical size during flight. The procedure described can be used to provide prognosis tools with the current damage configuration.

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.

Nonstructural damages of reinforced concrete buildings due to 2015 Ranau earthquake

NASA Astrophysics Data System (ADS)

Adiyanto, Mohd Irwan; Majid, Taksiah A.; Nazri, Fadzli Mohamed

2017-07-01

On 15th June 2016 a moderate earthquake with magnitude Mw5.9 was occurred in Sabah, Malaysia. Specifically, the epicentre was located at 16 km northwest of Ranau. Less than two days after the first event, a reconnaissance mission took action to investigate the damages on buildings. Since the reinforced concrete buildings in Ranau were designed based on gravity and wind load only, a lot of minor to severe damages was occurred. This paper presents the damages on the nonstructural elements of reinforced concrete buildings due to Ranau earthquake. The assessment was conducted via in-situ field investigation covering the visual observation, taking photo, and interview with local resident. Based on in-situ field investigation, there was a lot of damages occurred on the nonstructural elements like the brick walls. Such damages cannot be neglected since it can cause injury and fatality to the victims. Therefore, it can be concluded that the installation of nonstructural elements should be reviewed for the sake of safety.

Adaptive Finite Element Methods for Continuum Damage Modeling

NASA Technical Reports Server (NTRS)

Min, J. B.; Tworzydlo, W. W.; Xiques, K. E.

1995-01-01

The paper presents an application of adaptive finite element methods to the modeling of low-cycle continuum damage and life prediction of high-temperature components. The major objective is to provide automated and accurate modeling of damaged zones through adaptive mesh refinement and adaptive time-stepping methods. The damage modeling methodology is implemented in an usual way by embedding damage evolution in the transient nonlinear solution of elasto-viscoplastic deformation problems. This nonlinear boundary-value problem is discretized by adaptive finite element methods. The automated h-adaptive mesh refinements are driven by error indicators, based on selected principal variables in the problem (stresses, non-elastic strains, damage, etc.). In the time domain, adaptive time-stepping is used, combined with a predictor-corrector time marching algorithm. The time selection is controlled by required time accuracy. In order to take into account strong temperature dependency of material parameters, the nonlinear structural solution a coupled with thermal analyses (one-way coupling). Several test examples illustrate the importance and benefits of adaptive mesh refinements in accurate prediction of damage levels and failure time.

Multi-level damage identification with response reconstruction

NASA Astrophysics Data System (ADS)

Zhang, Chao-Dong; Xu, You-Lin

2017-10-01

Damage identification through finite element (FE) model updating usually forms an inverse problem. Solving the inverse identification problem for complex civil structures is very challenging since the dimension of potential damage parameters in a complex civil structure is often very large. Aside from enormous computation efforts needed in iterative updating, the ill-condition and non-global identifiability features of the inverse problem probably hinder the realization of model updating based damage identification for large civil structures. Following a divide-and-conquer strategy, a multi-level damage identification method is proposed in this paper. The entire structure is decomposed into several manageable substructures and each substructure is further condensed as a macro element using the component mode synthesis (CMS) technique. The damage identification is performed at two levels: the first is at macro element level to locate the potentially damaged region and the second is over the suspicious substructures to further locate as well as quantify the damage severity. In each level's identification, the damage searching space over which model updating is performed is notably narrowed down, not only reducing the computation amount but also increasing the damage identifiability. Besides, the Kalman filter-based response reconstruction is performed at the second level to reconstruct the response of the suspicious substructure for exact damage quantification. Numerical studies and laboratory tests are both conducted on a simply supported overhanging steel beam for conceptual verification. The results demonstrate that the proposed multi-level damage identification via response reconstruction does improve the identification accuracy of damage localization and quantization considerably.

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 study of a valveless pulse detonation rocket engine using nontoxic hypergolic propellants

NASA Astrophysics Data System (ADS)

Kan, Brandon K.

A pulsed detonation rocket engine concept was explored through the use of hypergolic propellants in a fuel-centered pintle injector combustor. The combustor design yielded a simple open ended chamber with a pintle type injection element and pressure instrumentation. High-frequency pressure measurements from the first test series showed the presence of large pressure oscillations in excess of 2000 psia at frequencies between 400-600 hz during operation. High-speed video confirmed the high-frequency pulsed behavior and large amounts of after burning. Damaged hardware and instrumentation failure limited the amount of data gathered in the first test series, but the experiments met original test objectives of producing large over-pressures in an open chamber. A second test series proceeded by replacing hardware and instrumentation, and new data showed that pulsed events produced under expanded exhaust prior to pulsing, peak pressures around 8000 psi, and operating frequencies between 400-800 hz. Later hot-fires produced no pulsed behavior despite undamaged hardware. The research succeeded in producing pulsed combustion behavior using hypergolic fuels in a pintle injector setup and provided insights into design concepts that would assist future injector designs and experimental test setups.

Detection of Matrix Crack Density of CFRP using an Electrical Potential Change Method with Multiple Probes

NASA Astrophysics Data System (ADS)

Todoroki, Akira; Omagari, Kazuomi

Carbon Fiber Reinforced Plastic (CFRP) laminates are adopted for fuel tank structures of next generation space rockets or automobiles. Matrix cracks may cause fuel leak or trigger fatigue damage. A monitoring system of the matrix crack density is required. The authors have developed an electrical resistance change method for the monitoring of delamination cracks in CFRP laminates. Reinforcement fibers are used as a self-sensing system. In the present study, the electric potential method is adopted for matrix crack density monitoring. Finite element analysis (FEA) was performed to investigate the possibility of monitoring matrix crack density using multiple electrodes mounted on a single surface of a specimen. The FEA reveals the matrix crack density increases electrical resistance for a target segment between electrodes. Experimental confirmation was also performed using cross-ply laminates. Eight electrodes were mounted on a single surface of a specimen using silver paste after polishing of the specimen surface with sandpaper. The two outermost electrodes applied electrical current, and the inner electrodes measured electric voltage changes. The slope of electrical resistance during reloading is revealed to be an appropriate index for the detection of matrix crack density.

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

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.

Characteristics of Creep Damage for 60Sn-40Pb Solder Material

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

Wei, Y.; Chow, C.L.; Fang, H.E.

This paper presents a viscoplasticity model taking into account the effects of change in grain or phase size and damage on the characterization of creep damage in 60Sn-40Pb solder. Based on the theory of damage mechanics, a two-scalar damage model is developed for isotropic materials by introducing the free energy equivalence principle. The damage evolution equations are derived in terms of the damage energy release rates. In addition, a failure criterion is developed based on the postulation that a material element is said to have ruptured when the total damage accumulated in the element reaches a critical value. The damagemore » coupled viscoplasticity model is discretized and coded in a general-purpose finite element program known as ABAQUS through its user-defined material subroutine UMAT. To illustrate the application of the model, several example cases are introduced to analyze, both numerically and experimentally, the tensile creep behaviors of the material at three stress levels. The model is then applied to predict the deformation of a notched specimen under monotonic tension at room temperature (22 C). The results demonstrate that the proposed model can successfully predict the viscoplastic behavior of the solder material.« less

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...

Climate Science and the Responsibilities of Fossil Fuel Companies for Climate Damages and Adaptation

NASA Astrophysics Data System (ADS)

Frumhoff, P. C.; Ekwurzel, B.

2017-12-01

Policymakers in several jurisdictions are now considering whether fossil fuel companies might bear some legal responsibility for climate damages and the costs of adaptation to climate change potentially traceable to the emissions from their marketed products. Here, we explore how scientific research, outreach and direct engagement with industry leaders and shareholders have informed and may continue to inform such developments. We present the results of new climate model research quantifying the contribution of carbon dioxide and methane emissions traced to individual fossil fuel companies to changes in global temperature and sea level; explore the impact of such research and outreach on both legal and broader societal consideration of company responsibility; and discuss the opportunities and challenges for scientists to engage in further work in this area.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-01

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Method for storing nuclear fuel in respositories

DOEpatents

Schweitzer, D.G.; Sastre, C.

A method for storing radioactive spent fuel in repositories containing polyphenyl or silicon oil as the storage medium is disclosed. Polyphenyls and silicon oils are non-corrosive and are not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

Requirements to the procedure and stages of innovative fuel development

NASA Astrophysics Data System (ADS)

Troyanov, V.; Zabudko, L.; Grachyov, A.; Zhdanova, O.

2016-04-01

According to the accepted current understanding under the nuclear fuel we will consider the assembled active zone unit (Fuel assembly) with its structural elements, fuel rods, pellet column, structural materials of fuel rods and fuel assemblies. The licensing process includes justification of safe application of the proposed modifications, including design-basis and experimental justification of the modified items under normal operating conditions and in violation of normal conditions, including accidents as well. Besides the justification of modified units itself, it is required to show the influence of modifications on the performance and safety of the other Reactor Unit’ and Nuclear Plant’ elements (e.g. burst can detection system, transportation and processing operations during fuel handling), as well as to justify the new standards of fuel storage etc. Finally, the modified fuel should comply with the applicable regulations, which often becomes a very difficult task, if only because those regulations, such as the NP-082-07, are not covered modification issues. Making amendments into regulations can be considered as the only solution, but the process is complicated and requires deep grounds for amendments. Some aspects of licensing new nuclear fuel are considered the example of mixed nitride uranium -plutonium fuel application for the BREST reactor unit.

Electric cartridge-type heater for producing a given non-uniform axial power distribution

DOEpatents

Clark, D.L.; Kress, T.S.

1975-10-14

An electric cartridge heater is provided to simulate a reactor fuel element for use in safety and thermal-hydraulic tests of model nuclear reactor systems. The electric heat-generating element of the cartridge heater consists of a specifically shaped strip of metal cut with variable width from a flat sheet of the element material. When spirally wrapped around a mandrel, the strip produces a coiled element of the desired length and diameter. The coiled element is particularly characterized by an electrical resistance that varies along its length due to variations in strip width. Thus, the cartridge heater is constructed such that it will produce a more realistic simulation of the actual nonuniform (approximately ''chopped'' cosine) power distribution of a reactor fuel element.

Develop an piezoelectric sensing based on SHM system for nuclear dry storage system

NASA Astrophysics Data System (ADS)

Ma, Linlin; Lin, Bin; Sun, Xiaoyi; Howden, Stephen; Yu, Lingyu

2016-04-01

In US, there are over 1482 dry cask storage system (DCSS) in use storing 57,807 fuel assemblies. Monitoring is necessary to determine and predict the degradation state of the systems and structures. Therefore, nondestructive monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health" for the safe operation of nuclear power plants (NPP) and radioactive waste storage systems (RWSS). Innovative approaches are desired to evaluate the degradation and damage of used fuel containers under extended storage. Structural health monitoring (SHM) is an emerging technology that uses in-situ sensory system to perform rapid nondestructive detection of structural damage as well as long-term integrity monitoring. It has been extensively studied in aerospace engineering over the past two decades. This paper presents the development of a SHM and damage detection methodology based on piezoelectric sensors technologies for steel canisters in nuclear dry cask storage system. Durability and survivability of piezoelectric sensors under temperature influence are first investigated in this work by evaluating sensor capacitance and electromechanical admittance. Toward damage detection, the PES are configured in pitch catch setup to transmit and receive guided waves in plate-like structures. When the inspected structure has damage such as a surface defect, the incident guided waves will be reflected or scattered resulting in changes in the wave measurements. Sparse array algorithm is developed and implemented using multiple sensors to image the structure. The sparse array algorithm is also evaluated at elevated temperature.

Modeling 3D PCMI using the Extended Finite Element Method with higher order elements

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

Jiang, W.; Spencer, Benjamin W.

2017-03-31

This report documents the recent development to enable XFEM to work with higher order elements. It also demonstrates the application of higher order (quadratic) elements to both 2D and 3D models of PCMI problems, where discrete fractures in the fuel are represented using XFEM. The modeling results demonstrate the ability of the higher order XFEM to accurately capture the effects of a crack on the response in the vicinity of the intersecting surfaces of cracked fuel and cladding, as well as represent smooth responses in the regions away from the crack.

Calculation of Free-Atom Fractions in Hydrocarbon-Fueled Rocket Engine Plume

NASA Technical Reports Server (NTRS)

Verma, Satyajit

2006-01-01

Free atom fractions (Beta) of nine elements are calculated in the exhaust plume of CH4- oxygen and RP-1-oxygen fueled rocket engines using free energy minimization method. The Chemical Equilibrium and Applications (CEA) computer program developed by the Glenn Research Center, NASA is used for this purpose. Data on variation of Beta in both fuels as a function of temperature (1600 K - 3100 K) and oxygen to fuel ratios (1.75 to 2.25 by weight) is presented in both tabular and graphical forms. Recommendation is made for the Beta value for a tenth element, Palladium. The CEA computer code was also run to compare with experimentally determined Beta values reported in literature for some of these elements. A reasonable agreement, within a factor of three, between the calculated and reported values is observed. Values reported in this work will be used as a first approximation for pilot rocket engine testing studies at the Stennis Space Center for at least six elements Al, Ca, Cr, Cu, Fe and Ni - until experimental values are generated. The current estimates will be improved when more complete thermodynamic data on the remaining four elements Ag, Co, Mn and Pd are added to the database. A critique of the CEA code is also included.

Reduced Toxicity Fuel Satellite Propulsion System Including Fuel Cell Reformer with Alcohols Such as Methanol

NASA Technical Reports Server (NTRS)

Schneider, Steven J. (Inventor)

2001-01-01

A reduced toxicity fuel satellite propulsion system including a reduced toxicity propellant supply for consumption in an axial class thruster and an ACS class thruster. The system includes suitable valves and conduits for supplying the reduced toxicity propellant to the ACS decomposing element of an ACS thruster. The ACS decomposing element is operative to decompose the reduced toxicity propellant into hot propulsive gases. In addition the system includes suitable valves and conduits for supplying the reduced toxicity propellant to an axial decomposing element of the axial thruster. The axial decomposing element is operative to decompose the reduced toxicity propellant into hot gases. The system further includes suitable valves and conduits for supplying a second propellant to a combustion chamber of the axial thruster, whereby the hot gases and the second propellant auto-ignite and begin the combustion process for producing thrust.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

2012-01-01

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

Metallic impurities-silicon carbide interaction in HTGR fuel particles

NASA Astrophysics Data System (ADS)

Minato, Kazuo; Ogawa, Toru; Kashimura, Satoru; Fukuda, Kousaku; Shimizu, Michio; Tayama, Yoshinobu; Takahashi, Ishio

1990-12-01

Corrosion of the coating layers of silicon carbide (SiC) by metallic impurities was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors with an optical microscope and an electron probe micro-analyzer. The SiC layers were attacked from the outside of the particles. The main element observed in the corroded areas was iron, but sometimes iron and nickel were found. These elements must have been contained as impurities in the graphite matrix in which the coated particles were dispersed. Since these elements are more stable thermodynamically in the presence of SiC than in the presence of graphite at irradiation temperatures, they were transferred to the SiC layer to form more stable silicides. During fuel manufacturing processes, intensive care should be taken to prevent the fuel from being contaminated with those elements which react with SiC.

Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar

2012-01-01

A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

KSC-2010-4843

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- Freedom Star, one of NASA's solid rocket booster retrieval ships, pulls the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4850

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- Freedom Star, one of NASA's solid rocket booster retrieval ships, pulls the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett

KSC-2010-4846

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- Freedom Star, one of NASA's solid rocket booster retrieval ships, pulls the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Jack Pfaller

KSC-2010-4830

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- Freedom Star, one of NASA's solid rocket booster retrieval ships, ushers the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4853

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- Freedom Star, one of NASA's solid rocket booster retrieval ships, pulls the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett

KSC-2010-4856

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- NASA's Pegasus barge moves through the bridge at Port Canaveral, Fla. The barge is carrying the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett

KSC-2010-4817

NASA Image and Video Library

2010-09-22

CAPE CANAVERAL, Fla. -- This view at dusk from the stern of Freedom Star, one of NASA's solid rocket booster retrieval ships, shows the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, as it is transported to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea before being offloaded and moved to Kennedy's Vehicle Assembly Building where it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4829

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- Freedom Star, one of NASA's solid rocket booster retrieval ships, ushers the Space Shuttle Program's last external fuel tank, ET-122, toward NASA's Kennedy Space Center in Florida. The tank traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans aboard the Pegasus Barge. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Frankie Martin

KSC-2010-4814

NASA Image and Video Library

2010-09-22

GULFPORT, La. -- This view from the captain's deck of Freedom Star, one of NASA's solid rocket booster retrieval ships, shows the Pegasus Barge carrying the Space Shuttle Program's last external fuel tank, ET-122, as it is escorted from Gulfport, La., to NASA's Kennedy Space Center in Florida. The tank will travel 900 miles by sea before being offloaded and moved to Kennedy's Vehicle Assembly Building. There it will be integrated to space shuttle Endeavour for the STS-134 mission to the International Space Station. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. STS-134, targeted to launch Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. Photo credit: NASA/Kim Shiflett

KSC-2010-4859

NASA Image and Video Library

2010-09-27

CAPE CANAVERAL, Fla. -- NASA's Pegasus barge is pulled toward NASA's Kennedy Space Center in Florida by a tug boat. The barge is carrying the Space Shuttle Program's last external fuel tank, ET-122 and traveled 900 miles by sea from NASA's Michoud Assembly Facility in New Orleans. After reaching the Turn Basin at Kennedy, the tank will be offloaded and moved to the Vehicle Assembly Building where it eventually will be attached to space shuttle Endeavour for the STS-134 mission to the International Space Station. STS-134, targeted to launch in Feb. 2011, currently is scheduled to be the last mission in the Space Shuttle Program. The tank, which is the largest element of the space shuttle stack, was damaged during Hurricane Katrina in August 2005 and restored to flight configuration by Lockheed Martin Space Systems Company employees. Photo credit: NASA/Kim Shiflett