Comparative Performance Assessment of 5kW-Class Solid Oxide Fuel Cell Engines Integrated With Single/Dual-Spool Turbochargers

DTIC Science & Technology

2011-01-01

Comparative Performance Assessment of 5kW-Class Solid Oxide Fuel Cell Engines Integrated with Single/Dual-Spool Turbochargers So-Ryeok Oh, Jing Sun... Turbochargers 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT... fundamental operating regime to the part load performance. Two different mechanical designs are assumed: dual shaft and single shaft as the compressor

Assessment of Turbulence-Chemistry Interaction Models in the National Combustion Code (NCC) - Part I

NASA Technical Reports Server (NTRS)

Wey, Thomas Changju; Liu, Nan-suey

2011-01-01

This paper describes the implementations of the linear-eddy model (LEM) and an Eulerian FDF/PDF model in the National Combustion Code (NCC) for the simulation of turbulent combustion. The impacts of these two models, along with the so called laminar chemistry model, are then illustrated via the preliminary results from two combustion systems: a nine-element gas fueled combustor and a single-element liquid fueled combustor.

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.

Pyrolysis result of polyethylene waste as fuel for solid oxide fuel cell with samarium doped-ceria (SDC)-carbonate as electrolyte

NASA Astrophysics Data System (ADS)

Syahputra, R. J. E.; Rahmawati, F.; Prameswari, A. P.; Saktian, R.

2017-02-01

In this research, the result of pyrolysis on polyethylene was used as fuel for a solid oxide fuel cell (SOFC). The pyrolysis result is a liquid which consists of hydrocarbon chains. According to GC-MS analysis, the hydrocarbons mainly consist of C7 to C20 hydrocarbon chain. Then, the liquid was applied to a single cell of NSDC-L | NSDC | NSDC-L. NSDC is a composite SDC (samarium doped-ceria) with sodium carbonate. Meanwhile, NSDC-L is a composite of NSDC with LiNiCuO (LNC). NSDC and LNC were analyzed by X-ray diffraction to understand their crystal structure. The result shows that presence of carbonate did not change the crystal structure of SDC. SEM EDX analysis for fuel cell before and after being loaded with polyethylene oil to get information of element diffusion to the electrolyte. Meanwhile, the conductivity properties were investigated through impedance measurement. The presence of carbonate even increases the electrical conductivity. The single cell test with the pyrolysis result of polyethylene at 300 - 600 °C, found that the highest power density is at 600 °C with the maximum power density of 0.14 mW/cm2 and open circuit voltage of 0.4 Volt. Elemental analysis at three point spots of single cell NDSC-L |NSDC|NSDC-L found that a migration of ions was occurred during fuel operation at 300 - 600 °C.

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.

Axisymmetric single shear element combustion instability experiment

NASA Technical Reports Server (NTRS)

Breisacher, Kevin J.

1993-01-01

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

Axisymmetric single shear element combustion instability experiment

NASA Technical Reports Server (NTRS)

Breisacher, Kevin J.

1993-01-01

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

A method for determining optimum phasing of a multiphase propulsion system for a single-stage vehicle with linearized inert weight

NASA Technical Reports Server (NTRS)

Martin, J. A.

1974-01-01

A general analytical treatment is presented of a single-stage vehicle with multiple propulsion phases. A closed-form solution for the cost and for the performance and a derivation of the optimal phasing of the propulsion are included. Linearized variations in the inert weight elements are included, and the function to be minimized can be selected. The derivation of optimal phasing results in a set of nonlinear algebraic equations for optimal fuel volumes, for which a solution method is outlined. Three specific example cases are analyzed: minimum gross lift-off weight, minimum inert weight, and a minimized general function for a two-phase vehicle. The results for the two-phase vehicle are applied to the dual-fuel rocket. Comparisons with single-fuel vehicles indicate that dual-fuel vehicles can have lower inert weight either by development of a dual-fuel engine or by parallel burning of separate engines from lift-off.

Effects of Fuel Spray Modeling on Combustion Instability Predictions in a Single-Element Lean Direct Injection (LDI) Gas Turbine Combustor

DTIC Science & Technology

2014-09-01

evaporation in the vicinity of the injector . Recently, Kim and Menon 9 applied the same approach to study the characteristics of longitudinal...phenomena that govern the occurrence of combustion instabilities. The experiments involve a single injector element in a longitudinal mode combustor with...well characterized inflow conditions and a choked nozzle exit condition. Varying parameters such as the length of the air plenum, the combustor length

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

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

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

Computational Study of Combustion Dynamics in a Single-Element Lean Direct Injection Gas Turbine Combustor

DTIC Science & Technology

2013-12-01

instabilities for different equivalence ratios and fuel injector locations. Comparisons of the computational and experimental results are carried out using...the fuel injector and swirler as the full geometry. The full geometry in Fig. 1 (b) is the same as the one that was used in the experiments. In Fig. 1...combustion instabilities in both the simulations and the experiments. Fuel injector (Detail B) sits in the converging-diverging section connecting the air

SHEATHED TUBE AND APPARATUS AND METHOD OF PRODUCTION THEREOF

DOEpatents

Ohlinger, L.A.

1959-08-18

A tubular fuel element covered inside and out by a unitary covering tube originally about twice its length and of small enough diameter to fit snugly inside the fuel tube is described. The covering tube is then reentrantly folded back by a pressure-die mechanism over both ends of the fuel tube and againsts outside until the folded back ends of the covering tube meet where they are welded in a single seam running circumferentially around the middle of the resulting assembly.

A Very Large Eddy Simulation of the Nonreacting Flow in a Single-Element Lean Direct Injection Combustor Using PRNS with a Nonlinear Subscale Model

NASA Technical Reports Server (NTRS)

Shih, Tsan-Hsing; Liu, Nan-Suey

2009-01-01

Very large eddy simulation (VLES) of the nonreacting turbulent flow in a single-element lean direct injection (LDI) combustor has been successfully performed via the approach known as the partially resolved numerical simulation (PRNS/VLES) using a nonlinear subscale model. The grid is the same as the one used in a previous RANS simulation, which was considered as too coarse for a traditional LES simulation. In this study, we first carry out a steady RANS simulation to provide the initial flow field for the subsequent PRNS/VLES simulation. We have also carried out an unsteady RANS (URANS) simulation for the purpose of comparing its results with that of the PRNS/VLES simulation. In addition, these calculated results are compared with the experimental data. The present effort has demonstrated that the PRNS/VLES approach, while using a RANS type of grid, is able to reveal the dynamically important, unsteady large-scale turbulent structures occurring in the flow field of a single-element LDI combustor. The interactions of these coherent structures play a critical role in the dispersion of the fuel, hence, the mixing between the fuel and the oxidizer in a combustor.

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

Luna, R. E.

This paper provides a simple model for estimating the release of respirable aerosols resulting from an attack on a spent fuel cask using a high energy density device (HEDD). Two primary experiments have provided data on potential releases from spent fuel casks under HEDD attack. Sandia National Laboratories (SNL) conducted the first in the early 1980's and the second was sponsored by Gessellshaft fur Anlagen- and Reaktorsicherheit (GRS) in Germany and conducted in France in 1994. Both used surrogate spent fuel assemblies in real casks. The SNL experiments used un-pressurized fuel pin assemblies in a single element cask while themore » GRS tests used pressurized fuel pin assemblies in a 9-element cask. Data from the two test programs is reasonably consistent, given the differences in the experiments, but the use of the test data for prediction of releases resulting from HEDD attack requires a method for accounting for the effects of pin pressurization release and the ratio of pin plenum gas release to cask free volume (VR). To account for the effects of VR and to link the two data sources, a simple model has been developed that uses both the SNL data and the GRS data as well as recent test data on aerosols produced in experiments with single pellets subjected to HEDD effects conducted under the aegis of the International Consortium's Working Group on Sabotage of Transport and Storage Casks (WGSTSC). (authors)« less

A Low NO(x) Lean-Direct Injection, Multipoint Integrated Module Combuster Concept for Advanced Aircraft Gas Turbines

NASA Technical Reports Server (NTRS)

Tacina, Robert; Wey, Changlie; Laing, Peter; Mansour, Adel

2002-01-01

A low NO(x) emissions combustor has been demonstrated in flame-tube tests. A multipoint, lean-direct injection concept was used. Configurations were tested that had 25- and 36- fuel injectors in the size of a conventional single fuel injector. An integrated-module approach was used for the construction where chemically etched laminates, diffusion bonded together, combine the fuel injectors, air swirlers and fuel manifold into a single element. Test conditions were inlet temperatures up to 810 K, inlet pressures up to 2760 kPa, and flame temperatures up to 2100 K. A correlation was developed relating the NO(x) emissions with the inlet temperature, inlet pressure, fuel-air ratio and pressure drop. Assuming that 10 percent of the combustion air would be used for liner cooling and using a hypothetical engine cycle, the NO(x) emissions using the correlation from flame-tube tests were estimated to be less than 20 percent of the 1996 ICAO standard.

CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

DOEpatents

Weems, S.J.

1963-09-24

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

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.

Durability test on irradiated rock-like oxide fuels

NASA Astrophysics Data System (ADS)

Kuramoto, K.; Nitani, N.; Yamashita, T.

2003-06-01

For a profitable use of Pu, Japan Atomic Energy Research Institute has been promoting researches for once-through type fuels. The strategy consists of stable rock-like oxide fuel fabrication in conventional fuel facilities followed by almost complete Pu burning in LWR and disposal of chemically stable spent fuel without further processing. Because leach rates of hazardous nuclides, such as TRU and β-emitters, that have long half-lives, are very important for the evaluation of geological safety, leaching tests in deionized water at 363 K were performed with reference to the MCC-1 method. Five irradiated fuel pellets, a single phase fuel of a yttria-stabilized zirconia (YSZ) containing UO 2 (U-YSZ), two fuels of U-YSZ particle dispersed in MgAl 2O 4 (SPI) or Al 2O 3 (COR) matrix, two homogeneous-blended fuels of U-YSZ and SPI or COR powders, were submitted to the tests. Stainless steel containers with Au coating and ethylene propylene diene monomer were used as leaching vessels and packing, respectively. The evaluated normalized leach rates of Zr, U and Pu were obviously lower than those of the other important elements and nuclides. Americium, Np and especially Y showed unexpectedly high evaluated normalized leach rates. The volatile elements, Cs and I, showed enhanced leaching within particle-dispersed type fuels because of crack formation around the particle.

CFD-Based Design of a Filming Injector for N+3 Combustors

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Mongia, Hukam; Lee, Phil

2016-01-01

An effort was undertaken to perform CFD analysis of fluid flow in Lean-Direct Injection (LDI) combustors with axial swirl-venturi elements coupled with a new fuel-filming injector design for next-generation N+3 combustors. The National Combustion Code (NCC) was used to perform non-reacting and two-phase reacting flow computations on a N+3 injector configuration, in a single-element and a five-element injector array. All computations were performed with a consistent approach towards mesh-generation, spray-, ignition- and kinetics-modeling with the NCC. Computational predictions of the aerodynamics of the injector were used to arrive at an optimal injector design that met effective area, aerodynamics, and fuel-air mixing criteria. LDI-3 emissions (EINOx, EICO and UHC) were compared with the previous generation LDI-2 combustor experimental data at representative engine cycle conditions.

The Single Crew Module Concept for Exploration

NASA Technical Reports Server (NTRS)

Chambliss, Joe

2012-01-01

Many concepts have been proposed for exploring space. In early 2010 presidential direction called for reconsidering the approach to address changes in exploration destinations, use of new technologies and development of new capabilities to support exploration of space. Considering the proposed new technology and capabilities that NASA was directed to pursue, the single crew module (SCM) concept for a more streamlined approach to the infrastructure and conduct of exploration missions was developed. The SCM concept combines many of the new promising technologies with a central concept of mission architectures that uses a single habitat module for all phases of an exploration mission. Integrating mission elements near Earth and fully fueling them prior to departure of the vicinity of Earth provides the capability of using the single habitat both in transit to an exploration destination and while exploring the destination. The concept employs the capability to return the habitat and interplanetary propulsion system to Earth vicinity so that those elements can be reused on subsequent exploration missions. This paper describes the SCM concept, provides a top level mass estimate for the elements needed and trades the concept against Many concepts have been proposed for exploring space. In early 2010 presidential direction called for reconsidering the approach to address changes in exploration destinations, use of new technologies and development of new capabilities to support exploration of space. Considering the proposed new technology and capabilities that NASA was directed to pursue, the single crew module (SCM) concept for a more streamlined approach to the infrastructure and conduct of exploration missions was developed. The SCM concept combines many of the new promising technologies with a central concept of mission architectures that uses a single habitat module for all phases of an exploration mission. Integrating mission elements near Earth and fully fueling them prior to departure of the vicinity of Earth provides the capability of using the single habitat both in transit to an exploration destination and while exploring the destination. The concept employs the capability to return the habitat and interplanetary propulsion system to Earth vicinity so that those elements can be reused on subsequent exploration missions. This paper describes the SCM concept, provides a top level mass estimate for the elements needed and trades the concept against Constellation approaches for Lunar, Near Earth Asteroid and Mars Surface missions.

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.

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.

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...

Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

NASA Astrophysics Data System (ADS)

Najeeb, Umair

This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

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.

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.

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

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

Nuclear reactor composite fuel assembly

DOEpatents

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

1980-01-01

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

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

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

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

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

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)

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

The Single Habitat Module Concept for Exploration - Mission Planning and Mass Estimates

NASA Technical Reports Server (NTRS)

Chambliss, Joe

2013-01-01

The Single Habitat Module (SHM) concept approach to the infrastructure and conduct of exploration missions combines many of the new promising technologies with a central concept of mission architectures that use a single habitat module for all phases of an exploration mission. Integrating mission elements near Earth and fully fueling them prior to departure of the vicinity of Earth provides the capability of using the single habitat both in transit to an exploration destination and while exploring the destination. The concept employs the capability to return the habitat and interplanetary propulsion system to Earth vicinity so that those elements can be reused on subsequent exploration missions. This paper provides a review of the SHM concept, the advantages it provides, trajectory assessments related to use of a high specific impulse space based propulsion system, advances in mission planning and new mass estimates.

NEUTRONIC REACTOR

DOEpatents

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

1958-09-01

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

Nuclear reactor

DOEpatents

Thomson, Wallace B.

2004-03-16

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

Neutronic fuel element fabrication

DOEpatents

Korton, George

2004-02-24

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

Nuclear reactor control

DOEpatents

Cawley, William E.; Warnick, Robert F.

1982-01-01

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

Photographic Combustion Characterization of LOX/Hydrocarbon Type Propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1980-01-01

The advantages and limitations of using high speed photography to identify potential combustion anomalies (pops, fuel freezing, reactive stream separation (RSS), carbon formation) were demonstrated. Combustion evaluation criteria were developed for evaluating, characterizing, and screening promising low cost propellant combination(s) and injector element(s) for long life, reusable engine systems. Carbon formation and RSS mechanisms and trends were identified by using high speed color photography at speeds up to 6000 frames/sec. Single element injectors were tested with LOX/RP-1, LOX/Propane, LOX/Methane and LOX/Ammonia propellants. Tests were conducted using seven separate injector elements. Five different conventionally machined elements were tested: OFO Triplet; Rectangular Unlike Doublet (RUD); Unlike Doublet (UD); Like on Lke Doublet (LOL-EDM); and Slit Triplet.

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...

Catalytic Combustion for Ultra-Low NOx Hydrogen Turbines

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

Etemad, Shahrokh; Baird, Benjamin; Alavandi, Sandeep

2011-06-30

Precision Combustion, Inc., (PCI) in close collaboration with Solar Turbines, Incorporated, has developed and demonstrated a combustion system for hydrogen fueled turbines that reduces NOx to low single digit level while maintaining or improving current levels of efficiency and eliminating emissions of carbon dioxide. Full scale Rich Catalytic Hydrogen (RCH1) injector was developed and successfully tested at Solar Turbines, Incorporated high pressure test facility demonstrating low single digit NOx emissions for hydrogen fuel in the range of 2200F-2750F. This development work was based on initial subscale development for faster turnaround and reduced cost. Subscale testing provided promising results for 42%more » and 52% H2 with NOx emissions of less than 2 ppm with improved flame stability. In addition, catalytic reactor element testing for substrate oxidation, thermal cyclic injector testing to simulate start-stop operation in a gas turbine environment, and steady state 15 atm. operation testing were performed successfully. The testing demonstrated stable and robust catalytic element component life for gas turbine conditions. The benefit of the catalytic hydrogen combustor technology includes capability of delivering near-zero NOx without costly post-combustion controls and without requirement for added sulfur control. In addition, reduced acoustics increase gas turbine component life. These advantages advances Department of Energy (DOE’s) objectives for achievement of low single digit NOx emissions, improvement in efficiency vs. postcombustion controls, fuel flexibility, a significant net reduction in Integrated Gasification Combined Cycle (IGCC) system net capital and operating costs, and a route to commercialization across the power generation field from micro turbines to industrial and utility turbines.« less

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.

Investigation of the Thermal Stability of Nd(x)Sc(y)Zr(1-x-y)O(2-δ) Materials Proposed for Inert Matrix Fuel Applications.

PubMed

Hayes, John R; Grosvenor, Andrew P; Saoudi, Mouna

2016-02-01

Inert matrix fuels (IMF) consist of transuranic elements (i.e., Pu, Am, Np, Cm) embedded in a neutron transparent (inert) matrix and can be used to "burn up" (transmute) these elements in current or Generation IV nuclear reactors. Yttria-stabilized zirconia has been extensively studied for IMF applications, but the low thermal conductivity of this material limits its usefulness. Other elements can be used to stabilize the cubic zirconia structure, and the thermal conductivity of the fuel can be increased through the use of a lighter stabilizing element. To this end, a series of Nd(x)Sc(y)Zr(1-x-y)O(2-δ) materials has been synthesized via a co-precipitation reaction and characterized by multiple techniques (Nd was used as a surrogate for Am). The long-range and local structures of these materials were studied using powder X-ray diffraction, scanning electron microscopy, and X-ray absorption spectroscopy. Additionally, the stability of these materials over a range of temperatures has been studied by annealing the materials at 1100 and 1400 °C. It was shown that the Nd(x)Sc(y)Zr(1-x-y)O(2-δ) materials maintained a single cubic phase upon annealing at high temperatures only when both Nd and Sc were present with y ≥ 0.10 and x + y > 0.15.

Means for supporting fuel elements in a nuclear reactor

DOEpatents

Andrews, Harry N.; Keller, Herbert W.

1980-01-01

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

Fuel assembly for nuclear reactors

DOEpatents

Creagan, Robert J.; Frisch, Erling

1977-01-01

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

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

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

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

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

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...

Polypropylene oil as fuel for solid oxide fuel cell with samarium doped-ceria (SDC)-carbonate as electrolyte

NASA Astrophysics Data System (ADS)

Syahputra, R. J. E.; Rahmawati, F.; Prameswari, A. P.; Saktian, R.

2017-03-01

The research focusses on converting polypropylene oil as pyrolysis product of polypropylene plastic into an electricity. The converter was a direct liquid fuel-solid oxide fuel cell (SOFC) with cerium oxide based material as electrolyte. The polypropylene vapor flowed into fuel cell, in the anode side and undergo oxidation reaction, meanwhile, the Oxygen in atmosphere reduced into oxygen ion at cathode. The fuel cell test was conducted at 400 - 600 °C. According to GC-MS analysis, the polypropylene oil consist of C8 to C27 hydrocarbon chain. The XRD analysis result shows that Na2CO3 did not change the crystal structure of SDC even increases the electrical conductivity. The maximum power density is 0.079 mW.cm-2 at 773 K. The open circuite voltage is 0.77 volt. Chemical stability test by analysing the single cell at before and after fuel cell test found that ionic migration occured during fuel cell operation. It is supported by the change of elemental composition in the point position of electrolyte and at the electrolyte-electrode interface

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.

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.

Quantification of rare earth elements using laser-induced breakdown spectroscopy

DOE PAGES

Martin, Madhavi; Martin, Rodger C.; Allman, Steve; ...

2015-10-21

In this paper, a study of the optical emission as a function of concentration of laser-ablated yttrium (Y) and of six rare earth elements, europium (Eu), gadolinium (Gd), lanthanum (La), praseodymium (Pr), neodymium (Nd), and samarium (Sm), has been evaluated using the laser-induced breakdown spectroscopy (LIBS) technique. Statistical methodology using multivariate analysis has been used to obtain the sampling errors, coefficient of regression, calibration, and cross-validation of measurements as they relate to the LIBS analysis in graphite-matrix pellets that were doped with elements at several concentrations. Each element (in oxide form) was mixed in the graphite matrix in percentages rangingmore » from 1% to 50% by weight and the LIBS spectra obtained for each composition as well as for pure oxide samples. Finally, a single pellet was mixed with all the elements in equal oxide masses to determine if we can identify the elemental peaks in a mixed pellet. This dataset is relevant for future application to studies of fission product content and distribution in irradiated nuclear fuels. These results demonstrate that LIBS technique is inherently well suited for the future challenge of in situ analysis of nuclear materials. Finally, these studies also show that LIBS spectral analysis using statistical methodology can provide quantitative results and suggest an approach in future to the far more challenging multielemental analysis of ~ 20 primary elements in high-burnup nuclear reactor fuel.« less

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)

Numerical Simulations of Two-Phase Reacting Flow in a Single-Element Lean Direct Injection (LDI) Combustor Using NCC

NASA Technical Reports Server (NTRS)

Liu, Nan-Suey; Shih, Tsan-Hsing; Wey, C. Thomas

2011-01-01

A series of numerical simulations of Jet-A spray reacting flow in a single-element lean direct injection (LDI) combustor have been conducted by using the National Combustion Code (NCC). The simulations have been carried out using the time filtered Navier-Stokes (TFNS) approach ranging from the steady Reynolds-averaged Navier-Stokes (RANS), unsteady RANS (URANS), to the dynamic flow structure simulation (DFS). The sub-grid model employed for turbulent mixing and combustion includes the well-mixed model, the linear eddy mixing (LEM) model, and the filtered mass density function (FDF/PDF) model. The starting condition of the injected liquid spray is specified via empirical droplet size correlation, and a five-species single-step global reduced mechanism is employed for fuel chemistry. All the calculations use the same grid whose resolution is of the RANS type. Comparisons of results from various models are presented.

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.

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.

Evidence for a single impact at the Cretaceous-Tertiary boundary from trace elements

NASA Technical Reports Server (NTRS)

Gilmour, Iain; Anders, Edward

1988-01-01

Not only meteoritic elements (Ir, Ni, Au, Pt metals), but also some patently non-meteoritic elements (As, Sb) are enriched at the K-T boundary. Eight enriched elements at 7 K-T sites were compared and it was found that: All have fairly constant proportions to Ir and Kilauea (invoked as an example of a volcanic source of Ir by opponents of the impact theory) has too little of 7 of these 8 elements to account for the boundary enrichments. The distribution of trace elements at the K-T boundary was reexamined using data from 11 sites for which comprehensive are available. The meteoritic component can be assessed by first normalizing the data to Ir, the most obviously extraterrestrial element, and then to Cl chondrites. The double normalization reduces the concentration range from 11 decades to 5 and also facilitates the identification of meteoritic elements. At sites where trace elements were analyzed in sub-divided samples of boundary clay, namely, Caravaca (SP), Stevns Klint (DK), Flaxbourne River (NZ) and Woodside Creek (NZ), Sb, As and Zn are well correlated with Ir across the boundary implying a common deposition mechanism. Elemental carbon is also enriched by up to 10,000 x in boundary clay from 5 K-T sides and is correlated with Ir across the boundary at Woodside Creek. While biomass would appear to be the primary fuel source for this carbon a contribution from a fossil fuel source may be necessary in order to account for the observed C abundance.

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

FCRD Advanced Reactor (Transmutation) Fuels Handbook

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

Janney, Dawn Elizabeth; Papesch, Cynthia Ann

2016-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, importantmore » to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, the handbook attempts to provide information about how well the property is known and how much variation exists between measurements. Although it includes some results from models, its primary focus is experimental data. The Handbook is organized in two sections: one with information about the U-Pu-Zr ternary and one with information about other elements and binary and vi ternary alloys in the U-Np-Pu-Am-La-Ce-Pr-Nd-Zr system. Within each section, information about elements is presented first, followed by information about binary alloys, then information about ternary alloys. The order in which the elements in each alloy are mentioned follows the order in the first sentence of this paragraph. Much of the information on the U-Pu-Zr system repeats information from the FCRD Transmutation Fuels Handbook 2015. Most of the other data has been published elsewhere (although scattered throughout numerous references, some quite obscure); however, some data from Idaho National Laboratory is presented here for the first time. As the FCRD programmatic mission evolves, future editions of this handbook will begin to include other advanced reactor fuel designs and compositions. Hence, the title of the handbook will transition to the Advanced Reactor Fuels Handbook.« 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)

Design considerations for a Space Shuttle Main Engine turbine blade made of single crystal material

NASA Technical Reports Server (NTRS)

Abdul-Aziz, A.; August, R.; Nagpal, V.

1993-01-01

Nonlinear finite-element structural analyses were performed on the first stage high-pressure fuel turbopump blade of the Space Shuttle Main Engine. The analyses examined the structural response and the dynamic characteristics at typical operating conditions. Single crystal material PWA-1480 was considered for the analyses. Structural response and the blade natural frequencies with respect to the crystal orientation were investigated. The analyses were conducted based on typical test stand engine cycle. Influence of combined thermal, aerodynamic, and centrifugal loadings was considered. Results obtained showed that the single crystal secondary orientation effects on the maximum principal stresses are not highly significant.

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.

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

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

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

1998-06-01

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

FUEL ELEMENT FOR NEUTRONIC REACTORS

DOEpatents

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

1961-01-17

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

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

Darrieus-Landau instability of premixed flames enhanced by fuel droplets

NASA Astrophysics Data System (ADS)

Nicoli, Colette; Haldenwang, Pierre; Denet, Bruno

2017-07-01

Recent experiments on spray flames propagating in a Wilson cloud chamber have established that spray flames are much more sensitive to wrinkles or corrugations than single-phase flames. To propose certain elements of explanation, we numerically study the Darrieus-Landau (or hydrodynamic) instability (DL-instability) developing in premixtures that contain an array of fuel droplets. Two approaches are compared: numerical simulation starting from the general conservation laws in reactive media, and the numerical computation of Sivashinsky-type model equations for DL-instability. Both approaches provide us with results in deep agreement. It is first shown that the presence of droplets in fuel-air premixtures induces initial perturbations which are large enough to trigger the DL-instability. Second, the droplets are responsible for additional wrinkles when the DL-instability is developed. The latter wrinkles are of length scales shorter than those of the DL-instability, in such a way that the DL-unstable spray flames have a larger front surface and therefore propagate faster than the single-phase ones when subjected to the same instability.

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.)

Tribology and energy efficiency: from molecules to lubricated contacts to complete machines.

PubMed

Taylor, Robert Ian

2012-01-01

The impact of lubricants on energy efficiency is considered. Molecular details of base oils used in lubricants can have a great impact on the lubricant's physical properties which will affect the energy efficiency performance of a lubricant. In addition, molecular details of lubricant additives can result in significant differences in measured friction coefficients for machine elements operating in the mixed/boundary lubrication regime. In single machine elements, these differences will result in lower friction losses, and for complete systems (such as cars, trucks, hydraulic circuits, industrial gearboxes etc.) lower fuel consumption or lower electricity consumption can result.

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)

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

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

E.T.; James P. Meagher; Prasad Apte

2002-12-31

This topical report summarizes work accomplished for the Program from November 1, 2001 to December 31, 2002 in the following task areas: Task 1: Materials Development; Task 2: Composite Development; Task 4: Reactor Design and Process Optimization; Task 8: Fuels and Engine Testing; 8.1 International Diesel Engine Program; 8.2 Nuvera Fuel Cell Program; and Task 10: Program Management. Major progress has been made towards developing high temperature, high performance, robust, oxygen transport elements. In addition, a novel reactor design has been proposed that co-produces hydrogen, lowers cost and improves system operability. Fuel and engine testing is progressing well, but wasmore » delayed somewhat due to the hiatus in program funding in 2002. The Nuvera fuel cell portion of the program was completed on schedule and delivered promising results regarding low emission fuels for transportation fuel cells. The evaluation of ultra-clean diesel fuels continues in single cylinder (SCTE) and multiple cylinder (MCTE) test rigs at International Truck and Engine. FT diesel and a BP oxygenate showed significant emissions reductions in comparison to baseline petroleum diesel fuels. Overall through the end of 2002 the program remains under budget, but behind schedule in some areas.« less

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.

Development of thermoplastic composite aircraft structures

NASA Technical Reports Server (NTRS)

Renieri, Michael P.; Burpo, Steven J.; Roundy, Lance M.; Todd, Stephanie A.; Kim, H. J.

1992-01-01

Efforts focused on the use of thermoplastic composite materials in the development of structural details associated with an advanced fighter fuselage section with applicability to transport design. In support of these designs, mechanics developments were conducted in two areas. First, a dissipative strain energy approach to material characterization and failure prediction, developed at the Naval Research Laboratory, was evaluated as a design/analysis tool. Second, a finite element formulation for thick composites was developed and incorporated into a lug analysis method which incorporates pin bending effects. Manufacturing concepts were developed for an upper fuel cell cover. A detailed trade study produced two promising concepts: fiber placement and single-step diaphragm forming. Based on the innovative design/manufacturing concepts for the fuselage section primary structure, elements were designed, fabricated, and structurally tested. These elements focused on key issues such as thick composite lugs and low cost forming of fastenerless, stiffener/moldine concepts. Manufacturing techniques included autoclave consolidation, single diaphragm consolidation (SDCC) and roll-forming.

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.)

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...

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

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

NASA Astrophysics Data System (ADS)

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

2010-10-01

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

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.

The Single Habitat Module Concept a Streamlined Way to Explore

NASA Technical Reports Server (NTRS)

Chambliss, Joe

2012-01-01

Many concepts have been proposed for exploring space. In early 2010 presidential direction called for reconsidering the approach to address changes in exploration destinations, use of new technologies and development of new capabilities to support exploration of space. Considering the proposed new technologies and capabilities that NASA was directed to pursue, the Single Habitathabitat module (SHMSHM) concept for a more streamlined approach to the infrastructure and conduct of exploration missions was developed. The SHM concept combines many of the new promising technologies with a central concept of mission architectures that uses a single habitat module for all phases of an exploration mission. Integrating mission elements near Earth and fully fueling them prior to departure of the vicinity of Earth provides the capability of using the single habitat both in transit to an exploration destination and while exploring the destination. The concept employs the capability to return the habitat and interplanetary propulsion system to Earth vicinity so that those elements can be reused on subsequent exploration missions. This paper describes the SHM concept, and the advantages it provides to accomplish exploration objectives.

The Single Crew Module Concept a Streamlined Way to Explore

NASA Technical Reports Server (NTRS)

Chambliss, Joe

2012-01-01

Many concepts have been proposed for exploring space. In early 2010 presidential direction called for reconsidering the approach to address changes in exploration destinations, use of new technologies and development of new capabilities to support exploration of space. Considering the proposed new technology and capabilities that NASA was directed to pursue, the single crew module (SCM) concept for a more streamlined approach to the infrastructure and conduct of exploration missions was developed. The SCM concept combines many of the new promising technologies with a central concept of mission architectures that uses a single habitat module for all phases of an exploration mission. Integrating mission elements near Earth and fully fueling them prior to departure of the vicinity of Earth provides the capability of using the single habitat both in transit to an exploration destination and while exploring the destination. The concept employs the capability to return the habitat and interplanetary propulsion system to Earth vicinity so that those elements can be reused on subsequent exploration missions. This paper describes the SCM concept, and the advantages it provides to accomplish exploration objectives.

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.

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

NASA Technical Reports Server (NTRS)

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

2005-01-01

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

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.

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.

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.

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.

EXAFS: New tool for study of battery and fuel cell materials

NASA Technical Reports Server (NTRS)

Mcbreen, James; Ogrady, William E.; Pandya, Kaumudi I.

1987-01-01

Extended X ray absorption fine structure (EXAFS) is a powerful technique for probing the local atomic structure of battery and fuel cell materials. The major advantages of EXAFS are that both the probe and the signal are X rays and the technique is element selective and applicable to all states of matter. This permits in situ studies of electrodes and determination of the structure of single components in composite electrodes, or even complete cells. EXAFS specifically probes short range order and yields coordination numbers, bond distances, and chemical identity of nearest neighbors. Thus, it is ideal for structural studies of ions in solution and the poorly crystallized materials that are often the active materials or catalysts in batteries and fuel cells. Studies on typical battery and fuel cell components are used to describe the technique and the capability of EXAFS as a structural tool in these applications. Typical experimental and data analysis procedures are outlined. The advantages and limitations of the technique are also briefly discussed.

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.

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.

Solid oxide fuel cell cathode with oxygen-reducing layer

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

Surdoval, Wayne A.; Berry, David A.; Shultz, Travis

The disclosure provides a SOFC comprised of an electrolyte, anode, and cathode, where the cathode comprises an MIEC and an oxygen-reducing layer. The oxygen-reducing layer is in contact with the MIEC, and the MIEC is generally between and separating the oxygen-reducing layer and the electrolyte. The oxygen-reducing layer is comprised of single element oxides, single element carbonates, or mixtures thereof, and has a thickness of less than about 30 nm. In a particular embodiment, the thickness is less than 5 nm. In another embodiment, the thickness is about 3 monolayers or less. The oxygen-reducing layer may be a continuous filmmore » or a discontinuous film with various coverage ratios. The oxygen-reducing layer at the thicknesses described may be generated on the MIEC surface using means known in the art such as, for example, ALD processes.« less

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/.

Advanced Fuels for Chemical Lasers

DTIC Science & Technology

1975-05-01

age that the analysis is not limited to a single distance from the injector. The four curves in Fig. 86 demonstrate the types of reactions that are...annular wall for this concentric-element-type injector. The rear manifold assembly is made of machined nickel 1late and contains the four - bolt-hole...fabrication k] and assembly of the fully cooled laser combustor components, a series of four tests was cordicted to make preliminary assessment of

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.

Predicting Large Deflections of Multiplate Fuel Elements Using a Monolithic FSI Approach

DOE PAGES

Curtis, Franklin G.; Freels, James D.; Ekici, Kivanc

2017-10-26

As part of the Global Threat Reduction Initiative, the Oak Ridge National Laboratory is evaluating conversion of fuel for the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium. Currently, multiphysics simulations that model fluid-structure interaction phenomena are being performed to ensure the safety of the reactor with the new fuel type. A monolithic solver that fully couples fluid and structural dynamics is used to model deflections in the new design. A classical experiment is chosen to validate the capabilities of the current solver and the method. Here, a single-plate simulation with various boundary conditions as well asmore » a five-plate simulation are presented. Finally, use of the monolithic solver provides stable solutions for the large deflections and the tight coupling of the fluid and structure and the maximum deflections are captured accurately.« less

Correlations Between Optical, Chemical and Physical Properties of Biomass Burn Aerosols

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

Hopkins, Rebecca J.; Lewis, Keith M.; Dessiaterik, Yury

2007-09-20

Single scattering albedo (ω) and Angstrom absorption coefficient (αap) values are measured at 405, 532 and 870 nm for aerosols generated during controlled laboratory combustion of twelve wildland fuels. Considerable fuel dependent variation in these optical properties is observed at these wavelengths. Complementary microspectroscopy techniques are used to elucidate spatially resolved local chemical bonding, carbon-to-oxygen atomic ratios, percent of sp2 hybridization (graphitic nature), elemental composition, particle size and morphology. These parameters are compared directly with the corresponding optical properties for each combustion product, facilitating an understanding of the fuel dependent variability observed. Results indicate that combustion products can be dividedmore » into three categories based on chemical, physical and optical properties. Only materials displaying a high degree of sp2 hybridization, with chemical and physical properties characteristic of ‘soot’ or black carbon, exhibit ω and αap values that indicate a high light absorbing capacity.« less

Development of molten carbonate fuel cells for power generation

NASA Astrophysics Data System (ADS)

1980-04-01

The broad and comprehensive program included elements of system definition, cell and system modeling, cell component development, cell testing in pure and contaminated environments, and the first stages of technology scale up. Single cells, with active areas of 45 sq cm and 582 sq cm, were operated at 650 C and improved to state of the art levels through the development of cell design concepts and improved electrolyte and electrode components. Performance was shown to degrade by the presence of fuel contaminants, such as sulfur and chlorine, and due to changes in electrode structure. Using conventional hot press fabrication techniques, electrolyte structures up to 20" x 20" were fabricated. Promising approaches were developed for nonhot pressed electrolyte structure fabrication and a promising electrolyte matrix material was identified. This program formed the basis for a long range effort to realize the benefits of molten carbonate fuel cell power plants.

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.

FUEL ELEMENT

DOEpatents

Bean, R.W.

1963-11-19

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

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

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

Rest, J.; Hofman, G.L.

1997-12-01

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

Reduced size fuel cell for portable applications

NASA Technical Reports Server (NTRS)

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

2004-01-01

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

Coolant mass flow equalizer for nuclear fuel

DOEpatents

Betten, Paul R.

1978-01-01

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

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

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

Advanced Turbulence Modeling Concepts

NASA Technical Reports Server (NTRS)

Shih, Tsan-Hsing

2005-01-01

The ZCET program developed at NASA Glenn Research Center is to study hydrogen/air injection concepts for aircraft gas turbine engines that meet conventional gas turbine performance levels and provide low levels of harmful NOx emissions. A CFD study for ZCET program has been successfully carried out. It uses the most recently enhanced National combustion code (NCC) to perform CFD simulations for two configurations of hydrogen fuel injectors (GRC- and Sandia-injector). The results can be used to assist experimental studies to provide quick mixing, low emission and high performance fuel injector designs. The work started with the configuration of the single-hole injector. The computational models were taken from the experimental designs. For example, the GRC single-hole injector consists of one air tube (0.78 inches long and 0.265 inches in diameter) and two hydrogen tubes (0.3 inches long and 0.0226 inches in diameter opposed at 180 degree). The hydrogen tubes are located 0.3 inches upstream from the exit of the air element (the inlet location for the combustor). To do the simulation, the single-hole injector is connected to a combustor model (8.16 inches long and 0.5 inches in diameter). The inlet conditions for air and hydrogen elements are defined according to actual experimental designs. Two crossing jets of hydrogen/air are simulated in detail in the injector. The cold flow, reacting flow, flame temperature, combustor pressure and possible flashback phenomena are studied. Two grid resolutions of the numerical model have been adopted. The first computational grid contains 0.52 million elements, the second one contains over 1.3 million elements. The CFD results have shown only about 5% difference between the two grid resolutions. Therefore, the CFD result obtained from the model of 1.3-million grid resolution can be considered as a grid independent numerical solution. Turbulence models built in NCC are consolidated and well tested. They can handle both coarse and fine grids near the wall. They can model the effect of anisotropy of turbulent stresses and the effect of swirling. The chemical reactions of Magnusson model and ILDM method were both used in this study.

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.

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.

Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

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

Chandler, David; Betzler, Ben; Hirtz, Gregory John

2016-09-01

The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se productionmore » capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.« less

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)

The Single Habitat Module Concept for Exploration - Mission Planning and Mass Estimates

NASA Technical Reports Server (NTRS)

Chambliss, Joe; Studak, J. W.

2013-01-01

The Single Habitat Module (SHM) concept approach to the infrastructure and conduct of exploration missions combines many of new promising technologies with a central concept of mission architectures that use a single habitat module for all phases of an exploration mission. Integrating mission elements near Earth and fully fueling them prior to departure of the vicinity of Earth provides the capability of using the single habitat both in transit to/from an exploration destination and while exploring the destination. The concept employs the capability to return the habitat and interplanetary propulsion system to Earth vicinity so that those elements can be reused on subsequent exploration missions. This paper provides an overview of the SHM concept and the advantages it provides. A summary of calculations of the mass of the habitat propulsion system (HPS) needed to get the habitat from Low Mars Orbit (LMO) to the surface and back to LMO and an overview of trajectory and mission mass assessments related to use of a high specific impulse space based propulsion system is provided. Those calculations lead to the conclusion that the SHM concept can significantly reduce the mass required and streamline mission operations to explore Mars (and thus all exploration destinations).

The Single Habitat Module Concept for Exploration - Mission Planning and Mass Estimates

NASA Technical Reports Server (NTRS)

Chambliss, Joe; Studak, J. W.

2013-01-01

The Single Habitat Module (SHM) concept approach to the infrastructure and conduct of exploration missions combines many new promising technologies with a central concept of mission architectures that use a single habitat module for all phases of an exploration mission. Integrating mission elements near Earth and fully fueling them prior to departure of the vicinity of Earth provides the capability of using the single habitat both in transit to/from an exploration destination and while exploring the destination. The concept employs the capability to return the habitat and interplanetary propulsion system to Earth vicinity so that those elements can be reused on subsequent exploration missions. This paper provides an overview of the SHM concept and the advantages it provides. The paper also provides a summary of calculations of the mass of the Habitat Propulsion System (HPS) needed to get the habitat from low-Mars orbit (LMO) to the surface and back to LMO, and an overview of trajectory and mission mass assessments related to use of a high specific impulse space-based propulsion system. Those calculations led to the conclusion that the SHM concept results in low total mass required and streamlines mission operations to explore Mars (or other exploration destinations).

Developing a laser shockwave model for characterizing diffusion bonded interfaces

NASA Astrophysics Data System (ADS)

Lacy, Jeffrey M.; Smith, James A.; Rabin, Barry H.

2015-03-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.

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

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

Bauer, Thomas L.; Krause, Michael G

1986-07-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2004-07-01

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

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

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

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

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

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

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

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

DOEpatents

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

1957-09-10

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

NUCLEAR REACTOR

DOEpatents

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

1958-12-01

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

JACKETED FISSIONABLE MEMBER

DOEpatents

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

1960-09-13

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

U-Mo Plate Blister Anneal Interim Report

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

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

2010-10-01

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

A 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.

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.

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.

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.

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

NASA Astrophysics Data System (ADS)

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

2015-06-01

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

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)

Fatigue Failure of Space Shuttle Main Engine Turbine Blades

NASA Technical Reports Server (NTRS)

Swanson, Gregrory R.; Arakere, Nagaraj K.

2000-01-01

Experimental validation of finite element modeling of single crystal turbine blades is presented. Experimental results from uniaxial high cycle fatigue (HCF) test specimens and full scale Space Shuttle Main Engine test firings with the High Pressure Fuel Turbopump Alternate Turbopump (HPFTP/AT) provide the data used for the validation. The conclusions show the significant contribution of the crystal orientation within the blade on the resulting life of the component, that the analysis can predict this variation, and that experimental testing demonstrates it.

FUEL ELEMENT CONSTRUCTION

DOEpatents

Simnad, M.T.

1961-08-15

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

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

DTIC Science & Technology

2007-04-01

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

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

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

2011-01-01

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

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

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

None

1960-10-31

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

Compact Fuel Element Environment Test

NASA Technical Reports Server (NTRS)

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

2012-01-01

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

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

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

D. D. Keiser; J. I. Cole

2007-09-01

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

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

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

Rest, J.; Hofman, G.L.

1997-06-01

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

IRRADIATION METHOD AND APPARATUS

DOEpatents

Cabell, C.P.

1962-12-18

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

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

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

1992-01-01

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

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

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

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

2008-07-15

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

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.

Propellant injection strategy for suppressing acoustic combustion instability

NASA Astrophysics Data System (ADS)

Diao, Qina

Shear-coaxial injector elements are often used in liquid-propellant-rocket thrust chambers, where combustion instabilities remain a significant problem. A conventional solution to the combustion instability problem relies on passive control techniques that use empirically-developed hardware such as acoustic baffles and tuned cavities. In addition to adding weight and decreasing engine performance, these devices are designed using trial-and-error methods, which do not provide the capability to predict the overall system stability characteristics in advance. In this thesis, two novel control strategies that are based on propellant fluid dynamics were investigated for mitigating acoustic instability involving shear-coaxial injector elements. The new control strategies would use a set of controlled injectors allowing local adjustment of propellant flow patterns for each operating condition, particularly when instability could become a problem. One strategy relies on reducing the oxidizer-fuel density gradient by blending heavier methane with the main fuel, hydrogen. Another strategy utilizes modifying the equivalence ratio to affect the acoustic impedance through mixing and reaction rate changes. The potential effectiveness of these strategies was assessed by conducting unit-physics experiments. Two different model combustors, one simulating a single-element injector test and the other a double-element injector test, were designed and tested for flame-acoustic interaction. For these experiments, the Reynolds number of the central oxygen jet was kept between 4700 and 5500 making the injector flames sufficiently turbulent. A compression driver, mounted on one side of the combustor wall, provided controlled acoustic excitation to the injector flames, simulating the initial phase of flame-acoustic interaction. Acoustic excitation was applied either as band-limited white noise forcing between 100 Hz and 5000 Hz or as single-frequency, fixed-amplitude forcing at 1150 Hz which represented a frequency least amplified by any resonance. Effects of each control strategy on flame-acoustic interaction were assessed in terms of modifying the acoustic resonance characteristics subject to white-noise excitation and changes in flame brush thickness under single-frequency excitation. In the methane blending experiments, the methane mole fraction was varied between 0% and 63%. Under white noise excitation, up to 16% shift in a resonant frequency was observed but the acoustic pressure spectra remained qualitatively similar. For the fixed frequency forcing, the spatial extent of flame-acoustic interaction was substantially reduced. In the other experiments, the equivalence ratio of the control injector was varied between zero and infinity, causing up to 40% shift in a resonant frequency as well as changes in the acoustic pressure spectrum. These results open up the possibility of employing flow-based control to prevent combustion instabilities in liquid-fueled rockets.

Wind tunnel investigation of a high lift system with pneumatic flow control

NASA Astrophysics Data System (ADS)

Victor, Pricop Mihai; Mircea, Boscoianu; Daniel-Eugeniu, Crunteanu

2016-06-01

Next generation passenger aircrafts require more efficient high lift systems under size and mass constraints, to achieve more fuel efficiency. This can be obtained in various ways: to improve/maintain aerodynamic performance while simplifying the mechanical design of the high lift system going to a single slotted flap, to maintain complexity and improve the aerodynamics even more, etc. Laminar wings have less efficient leading edge high lift systems if any, requiring more performance from the trailing edge flap. Pulsed blowing active flow control (AFC) in the gap of single element flap is investigated for a relatively large model. A wind tunnel model, test campaign and results and conclusion are presented.

Assessment of Reduced-Kinetics Mechanisms for Combustion of Jet Fuel in CFD Applications

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Kundu, Krihna P.; Yungster, Shaye J.

2014-01-01

A computational effort was undertaken to analyze the details of fluid flow in Lean-Direct Injection (LDI) combustors for next-generation LDI design. The National Combustor Code (NCC) was used to perform reacting flow computations on single-element LDI injector configurations. The feasibility of using a reduced chemical-kinetics approach, which optimizes the reaction rates and species to model the emissions characteristics typical of lean-burning gas-turbine combustors, was assessed. The assessments were performed with Reynolds- Averaged Navier-Stokes (RANS) and Time-Filtered Navier Stokes (TFNS) time-integration, with a Lagrangian spray model with the NCC code. The NCC predictions for EINOx and combustor exit temperature were compared with experimental data for two different single-element LDI injector configurations, with 60deg and 45deg axially swept swirler vanes. The effects of turbulence-chemistry interaction on the predicted flow in a typical LDI combustor were studied with detailed comparisons of NCC TFNS with experimental data.

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.

Developing a laser shockwave model for characterizing diffusion bonded interfaces

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

Lacy, Jeffrey M., E-mail: Jeffrey.Lacy@inl.gov; Smith, James A., E-mail: Jeffrey.Lacy@inl.gov; Rabin, Barry H., E-mail: Jeffrey.Lacy@inl.gov

2015-03-31

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengthsmore » in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.« less

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.

Design and characterization of a microbial fuel cell for the conversion of a lignocellulosic crop residue to electricity.

PubMed

Gregoire, K P; Becker, J G

2012-09-01

Agricultural crop residues contain high amounts of biochemical energy as cellulose and lignin. A portion of this biomass could be sustainably harvested for conversion to bioenergy to help offset fossil fuel consumption. In this study, the potential for converting lignocellulosic biomass directly to electricity in a microbial fuel cell (MFC) was explored. Design elements of tubular air cathode MFCs and leach-bed bioreactors were integrated to develop a new solid-substrate MFC in which cellulose hydrolysis, fermentation, and anode respiration occurred in a single chamber. Electricity was produced continuously from untreated corncob pellets for >60 d. Addition of rumen fluid increased power production, presumably by providing growth factors to anode-respiring bacteria. Periodic exposure to oxygen also increased power production, presumably by limiting the diversion of electrons to methanogenesis. In the absence of methanogenesis, bioaugmentation with Geobacter metallireducens further improved MFC performance. Under these conditions, the maximum power density was 230 mW/m(3). Copyright © 2012 Elsevier Ltd. All rights reserved.

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.

Parametric Design of Injectors for LDI-3 Combustors

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Mongia, Hukam; Lee, Phil

2015-01-01

Application of a partially calibrated National Combustion Code (NCC) for providing guidance in the design of the 3rd generation of the Lean-Direct Injection (LDI) multi-element combustion configuration (LDI-3) is summarized. NCC was used to perform non-reacting and two-phase reacting flow computations on several LDI-3 injector configurations in a single-element and a five-element injector array. All computations were performed with a consistent approach for mesh-generation, turbulence, spray simulations, ignition and chemical kinetics-modeling. Both qualitative and quantitative assessment of the computed flowfield characteristics of the several design options led to selection of an optimal injector LDI- 3 design that met all the requirements including effective area, aerodynamics and fuel-air mixing criteria. Computed LDI-3 emissions (namely, NOx, CO and UHC) will be compared with the prior generation LDI- 2 combustor experimental data at relevant engine cycle conditions.

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)

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.

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

CFD Based Design of a Filming Injector for N+3 Combustors

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Mongia, Hukam; Lee, Phil

2016-01-01

An effort was undertaken to perform CFD analysis of fluid flow in Lean-Direct Injection (LDI) combustors with axial swirl-venturi elements for next-generation LDI-3 combustor design. The National Combustion Code (NCC) was used to perform non-reacting and two-phase reacting flow computations for a newly-designed pre-filming type fuel injector LDI-3 injector, in a single-injector and a five-injector array configuration. All computations were performed with a consistent approach of mesh-optimization, spray-modeling, ignition and kinetics-modeling. Computational predictions of the aerodynamics of the single-injector were used to arrive at an optimized main-injector design that meets effective area and fuel-air mixing criteria. Emissions (EINOx) characteristics were predicted for a medium-power engine cycle condition, and will be compared with data when it is made available from experimental measurements. The use of a PDF-like turbulence-chemistry interaction model with NCC's Time-Filtered Navier-Stokes (TFNS) solver is shown to produce a significant impact on the CFD results, when compared with a laminar-chemistry TFNS approach for the five-injector computations.

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

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

NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

DOEpatents

Brooks, H.

1960-04-26

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

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

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

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

Ion chromatographic determination of sulfur in fuels

NASA Technical Reports Server (NTRS)

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

1978-01-01

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

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

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

Rest, J.; Hofman, G.L.

1999-04-01

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

Oil spill source identification by principal component analysis of electrospray ionization Fourier transform ion cyclotron resonance mass spectra.

PubMed

Corilo, Yuri E; Podgorski, David C; McKenna, Amy M; Lemkau, Karin L; Reddy, Christopher M; Marshall, Alan G; Rodgers, Ryan P

2013-10-01

One fundamental challenge with either acute or chronic oil spills is to identify the source, especially in highly polluted areas, near natural oil seeps, when the source contains more than one petroleum product or when extensive weathering has occurred. Here we focus on heavy fuel oil that spilled (~200,000 L) from two suspected fuel tanks that were ruptured on the motor vessel (M/V) Cosco Busan when it struck the San Francisco-Oakland Bay Bridge in November 2007. We highlight the utility of principal component analysis (PCA) of elemental composition data obtained by high resolution FT-ICR mass spectrometry to correctly identify the source of environmental contamination caused by the unintended release of heavy fuel oil (HFO). Using ultrahigh resolution electrospray ionization (ESI) Fourier transform ion cyclotron resonance mass spectrometry, we uniquely assigned thousands of elemental compositions of heteroatom-containing species in neat samples from both tanks and then applied principal component analysis. The components were based on double bond equivalents for constituents of elemental composition, CcHhN1S1. To determine if the fidelity of our source identification was affected by weathering, field samples were collected at various intervals up to two years after the spill. We are able to identify a suite of polar petroleum markers that are environmentally persistent, enabling us to confidently identify that only one tank was the source of the spilled oil: in fact, a single principal component could account for 98% of the variance. Although identification is unaffected by the presence of higher polarity, petrogenic oxidation (weathering) products, future studies may require removal of such species by anion exchange chromatography prior to mass spectral analysis due to their preferential ionization by ESI.

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)

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

Data evaluation of trace elements determined in Nigerian coal using cluster procedures.

PubMed

Ewa, I O B

2004-05-01

Large data-sets of elements determined by instrumental neutron activation analysis (INAA) require meaningful interpretation in order to determine the pattern of their existence in host matrices. This could be achieved using cluster procedures. Element abundances (Al, As, Ba, Br, Ca, Ce, Cs, Dy, Eu, Fe, Ga, Gd, Hf, K, La, Lu, Mg, Mn, Na, O, Rb, Sb, Sc, Sm, Sr, Ta, Tb, Th, Ti, U, V, Yb, Zn and Zr) of prepared and run-of-mine coals from eight principal mines (Onyeama, Ogbete, Enugu, Gombe, Asaba-Ugwashi, Okaba, Afikpo and Lafia ) in Nigeria were determined by INAA. Quality control of the measurements was assured by the re-determination of a standard reference material, NIST 1632a. These data-sets were then tested for multi-variate statistics using METHOD = SINGLE in the cluster procedure. The computer-assisted package SAS was used to generate the dendrograms while the algorithm used was stored Euclidean distances. The results showed a recognition pattern, useful for the interpretation of coalification histories and the prediction of fuel ranking for Nigerian coals. High segregation of coal fly ash was observed, while metallurgical coal grouped together with high-ranking coals of Okaba, Enugu and Obi (Lafia). Further work revealed some of these coals as having high gross calorific value (7908 kcal kg(-1) for Enugu coal; 7200 kcal kg(-1) for Okaba) and low sulphur thereby making them efficient fuel materials.

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.

Three dimensional CFD modeling and experimental validation of a single chamber solid oxide fuel cell fed by methane

NASA Astrophysics Data System (ADS)

Nguyen, H. T.; Le, M. V.; Nguyen, T. A.; Nguyen, T. A. N.

2017-06-01

The solid oxide fuel cell is one of the promising technologies for future energy demand. Solid oxide fuel cell operated in the single-chamber mode exhibits several advantages over conventional single oxide fuel cell due to the simplified, compact, sealing-free cell structure. There are some studies on simulating the behavior of this type of fuel cell but they mainly focus on the 2D model. In the present study, a three-dimensional numerical model of a single chamber solid oxide fuel cell (SOFC) is reported and solved using COMSOL Multiphysics software. Experiments of a planar button solid oxide fuel cell were used to verify the simulation results. The system is fed by methane and oxygen and operated at 700°C. The cathode is LSCF6482, the anode is GDC-Ni, the electrolyte is LDM and the operating pressure is 1 atm. There was a good agreement between the cell temperature and current voltage estimated from the model and measured from the experiment. The results indicate that the model is applicable for the single chamber solid oxide fuel cell and it can provide a basic for the design, scale up of single chamber solid oxide fuel cell system.

Method for recovering catalytic elements from fuel cell membrane electrode assemblies

DOEpatents

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

2012-06-26

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

Combined catalysts for the combustion of fuel in gas turbines

DOEpatents

Anoshkina, Elvira V.; Laster, Walter R.

2012-11-13

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

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.

An experimental study of gaseous exhaust emissions of diesel engine using blend of natural fatty acid methyl ester

NASA Astrophysics Data System (ADS)

Sudrajad, Agung; Ali, Ismail; Samo, Khalid; Faturachman, Danny

2012-09-01

Vegetable oil form in Natural Fatty Acid Methyl Ester (FAME) has their own advantages: first of all they are available everywhere in the world. Secondly, they are renewable as the vegetables which produce oil seeds can be planted year after year. Thirdly, they are friendly with our environment, as they seldom contain sulphur element in them. This makes vegetable fuel studies become current among the various popular investigations. This study is attempt to optimization of using blend FAME on diesel engine by experimental laboratory. The investigation experimental project is comparison between using blend FAME and base diesel fuel. The engine experiment is conducted with YANMAR TF120M single cylinder four stroke diesel engine set-up at variable engine speed with constant load. The data have been taken at each point of engine speed during the stabilized engine-operating regime. Measurement of emissions parameters at difference engine speed conditions have generally indicated lower in emission NOx, but slightly higher on CO2 emission. The result also shown that the blends FAME are good in fuel consumption and potentially good substitute fuels for diesel engine

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.

Space power reactor in-core thermionic multicell evolutionary (S-prime) design

NASA Astrophysics Data System (ADS)

Determan, William R.; Van Hagan, Tom H.

1993-01-01

A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

NUCLEAR REACTOR CORE

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

Bell, F.R.

1963-02-01

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

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

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

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

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

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

PubMed

Saqib, Naeem; Bäckström, Mattias

2014-12-01

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

Effect of Crystal Orientation on Fatigue Failure of Single Crystal Nickel Base Turbine Blade Superalloys

NASA Technical Reports Server (NTRS)

Arakere, N. K.; Swanson, G.

2002-01-01

High cycle fatigue (HCF) induced failures in aircraft gas turbine and rocket engine turbopump blades is a pervasive problem. Single crystal nickel turbine blades are being utilized in rocket engine turbopumps and jet engines throughout industry because of their superior creep, stress rupture, melt resistance, and thermomechanical fatigue capabilities over polycrystalline alloys. Currently the most widely used single crystal turbine blade superalloys are PWA 1480/1493, PWA 1484, RENE' N-5 and CMSX-4. These alloys play an important role in commercial, military and space propulsion systems. Single crystal materials have highly orthotropic properties making the position of the crystal lattice relative to the part geometry a significant factor in the overall analysis. The failure modes of single crystal turbine blades are complicated to predict due to the material orthotropy and variations in crystal orientations. Fatigue life estimation of single crystal turbine blades represents an important aspect of durability assessment. It is therefore of practical interest to develop effective fatigue failure criteria for single crystal nickel alloys and to investigate the effects of variation of primary and secondary crystal orientation on fatigue life. A fatigue failure criterion based on the maximum shear stress amplitude /Delta(sub tau)(sub max))] on the 24 octahedral and 6 cube slip systems, is presented for single crystal nickel superalloys (FCC crystal). This criterion reduces the scatter in uniaxial LCF test data considerably for PWA 1493 at 1200 F in air. Additionally, single crystal turbine blades used in the alternate advanced high-pressure fuel turbopump (AHPFTP/AT) are modeled using a large-scale three-dimensional finite element model. This finite element model is capable of accounting for material orthotrophy and variation in primary and secondary crystal orientation. Effects of variation in crystal orientation on blade stress response are studied based on 297 finite element model runs. Fatigue lives at critical points in the blade are computed using finite element stress results and the failure criterion developed. Stress analysis results in the blade attachment region are also presented. Results presented demonstrates that control of secondary and primary crystallographic orientation has the potential to significantly increase a component S resistance to fatigue crack growth with- out adding additional weight or cost. [DOI: 10.1115/1.1413767

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.

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...

Spatially-Resolved Analyses of Aerodynamic Fallout from a Uranium-Fueled Nuclear Test

DOE PAGES

Lewis, L. A.; Knight, K. B.; Matzel, J. E.; ...

2015-07-28

The fiive silicate fallout glass spherules produced in a uranium-fueled, near-surface nuclear test were characterized by secondary ion mass spectrometry, electron probe microanalysis, autoradiography, scanning electron microscopy, and energy-dispersive x-ray spectroscopy. Several samples display compositional heterogeneity suggestive of incomplete mixing between major elements and natural U ( 238U/ 235U = 0.00725) and enriched U. Samples exhibit extreme spatial heterogeneity in U isotopic composition with 0.02 < 235U/ 238U < 11.84 among all five spherules and 0.02 < 235U/ 238U < 7.41 within a single spherule. Moreover, in two spherules, the 235U/ 238U ratio is correlated with changes in major elementmore » composition, suggesting the agglomeration of chemically and isotopically distinct molten precursors. Two samples are nearly homogenous with respect to major element and uranium isotopic composition, suggesting extensive mixing possibly due to experiencing higher temperatures or residing longer in the fireball. Linear correlations between 234U/ 238U, 235U/ 238U, and 236U/ 238U ratios are consistent with a two-component mixing model, which is used to illustrate the extent of mixing between natural and enriched U end members.« less

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

Optimal dual-fuel propulsion for minimum inert weight or minimum fuel cost

NASA Technical Reports Server (NTRS)

Martin, J. A.

1973-01-01

An analytical investigation of single-stage vehicles with multiple propulsion phases has been conducted with the phasing optimized to minimize a general cost function. Some results are presented for linearized sizing relationships which indicate that single-stage-to-orbit, dual-fuel rocket vehicles can have lower inert weight than similar single-fuel rocket vehicles and that the advantage of dual-fuel vehicles can be increased if a dual-fuel engine is developed. The results also indicate that the optimum split can vary considerably with the choice of cost function to be minimized.

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

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.

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

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

Anthony L. Alberti; Todd S. Palmer; Javier Ortensi

2016-05-01

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately modelmore » the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.« less

Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

NASA Technical Reports Server (NTRS)

Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

1974-01-01

The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

LCRE and SNAP 50-DR-1 programs. Engineering progress report, January 1, 1963--March 31, 1963

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, primary coolant circuit, aaxiliary systems, fuel elements, instrumentation, materials development, and fabrication; and SNAP-50DR-1 specifications, fuel elements, pumps, steam generator, and materials development. (DCC)

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT ...

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

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT FRAME. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-4. INEL INDEX CODE NUMBER: 075 0701 60 851 151978. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

High temperature ceramic composition for hydrogen retention

DOEpatents

Webb, R.W.

1974-01-01

A ceramic coating for H retention in fuel elements is described. The coating has relatively low thermal neutron cross section, is not readily reduced by H at 1500 deg F, is adherent to the fuel element base metal, and is stable at reactor operating temperatures. (JRD)

JACKETED URANIUM SLUG

DOEpatents

Ohlinger, L.A.; Cooper, C.M.

1958-10-01

Fuel elements for nuclear reactors are described. Eacb fuel element is comprised of a solid cylindrical slug containing fissionable material enclosed within a fluid tight jacket of neutron permeable material such as aluminum. The jacket is provided with a flexible end cap and with a sealing member having a substantially fluid-tight fit within the jacket in tight abutment with the end cap and the end of the slug. A fluid passage is provided between the end of the slug and the cap whereby leakage fiuid is principally directed to the end of the slug. In this manner, any reaction between the fissionable material and fiuid which may take place occurs more rapidly at the end of the slug than along the sides between the slug and the jacket, thereby causing longitudinal expansion of the fuel element prior to radial expansion. The longitudinal expansion can be readily detected and the fuel element removed from the coolant tube before radial expansion causes it to become jammed in the tube.

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

Summary of space nuclear reactor power systems, 1983--1992

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

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

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.

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.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

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

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

SHAPED FISSIONABLE METAL BODIES

DOEpatents

Wigner, E.P.; Williamson, R.R.; Young, G.J.

1958-10-14

A technique is presented for grooving the surface of fissionable fuel elements so that expansion can take place without damage to the interior structure of the fuel element. The fissionable body tends to develop internal stressing when it is heated internally by the operation of the nuclear reactor and at the same time is subjected to surface cooling by the circulating coolant. By producing a grooved or waffle-like surface texture, the annular lines of tension stress are disrupted at equally spaced intervals by the grooves, thereby relieving the tension stresses in the outer portions of the body while also facilitating the removal of accumulated heat from the interior portion of the fuel element.

FUEL ELEMENT INTERLOCKING ARRANGEMENT

DOEpatents

Fortescue, P.; Nicoll, D.

1963-01-01

This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.; Kopelman, B.; Hausner, H.H.

1963-07-01

A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Hauth, J.J.; Anicetti, R.J.

1962-12-01

A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

FOIL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Noland, R.A.; Walker, D.E.; Spinrad, B.I.

1963-07-16

A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

The EPA National Fuels Surveillance Network. I. Trace constituents in gasoline and commercial gasoline fuel additives.

PubMed Central

Jungers, R H; Lee, R E; von Lehmden, D J

1975-01-01

A National Fuels Surveillance Network has been established to collect gasoline and other fuels through the 10 regional offices of the Environmental Protection Agency. Physical, chemical, and trace element analytical determinations are made on the collected fuel samples to detect components which may present an air pollution hazard or poison exhaust catalytic control devices. A summary of trace elemental constituents in over 50 gasoline samples and 18 commercially marketed consumer purchased gasoline additives is presented. Quantities of Mn, Ni, Cr, Zn, Cu, Fe, Sb, B, Mg, Pb, and S were found in most regular and premium gasoline. Environmental implications of trace constituents in gasoline are discussed. PMID:1157783

Multiple fuel supply system for an internal combustion engine

DOEpatents

Crothers, William T.

1977-01-01

A multiple fuel supply or an internal combustion engine wherein phase separation of components is deliberately induced. The resulting separation permits the use of a single fuel tank to supply components of either or both phases to the engine. Specifically, phase separation of a gasoline/methanol blend is induced by the addition of a minor amount of water sufficient to guarantee separation into an upper gasoline phase and a lower methanol/water phase. A single fuel tank holds the two-phase liquid with separate fuel pickups and separate level indicators for each phase. Either gasoline or methanol, or both, can be supplied to the engine as required by predetermined parameters. A fuel supply system for a phase-separated multiple fuel supply contained in a single fuel tank is described.

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.

SODIUM DEUTERIUM REACTOR

DOEpatents

Oppenheimer, E.D.; Weisberg, R.A.

1963-02-26

This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

2011-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

Axially staggered seed-blanket reactor fuel module construction

DOEpatents

Cowell, Gary K.; DiGuiseppe, Carl P.

1985-01-01

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

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-07-11

Nuclear reactor fuel elements of the type in which the flssionsble material is in ceramic form, such as uranium dioxide, are described. The fuel element is comprised of elongated inner and outer concentric spaced tubular members providing an annular space therebetween for receiving the fissionable material, the annular space being closed at both ends and the inner tube being open at both ends. The fuel is in the form of compressed pellets of ceramic fissionsble material having the configuration of split bushings formed with wedge surfaces and arranged in seriated inner and outer concentric groups which are urged against the respective tubes in response to relative axial movement of the pellets in the direction toward each other. The pairs of pellets are axially urged together by a resilient means also enclosed within the annulus. This arrangement-permits relative axial displacement of the pellets during use dial stresses on the inner and outer tube members and yet maintains the fuel pellets in good thermal conductive relationship therewith.

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.

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.

Space storable propellant performance program coaxial injector characterization

NASA Technical Reports Server (NTRS)

Burick, R. J.

1972-01-01

An experimental program was conducted to characterize the circular coaxial injector concept for application with the space-storable gas/liquid propellant combination FLOX(82.6% F2)/CH4(g) at high pressure. The primary goal of the program was to obtain high characteristic velocity efficiency in conjunction with acceptable injector/chamber compatibility. A series of subscale (single element) cold flow and hot fire experiments was employed to establish design criteria for a 3000-lbf (sea level) engine operating at 500 psia. The subscale experiments characterized both high performance core elements and peripheral elements with enhanced injector/chamber compatibility. The full-scale injector which evolved from the study demonstrated a performance level of 99 percent of the theoretical shifting characteristic exhaust velocity with low chamber heat flux levels. A 44-second-duration firing demonstrated the durability of the injector. Parametric data are presented that are applicable for the design of circular, coaxial injectors that operate with injection dynamics (fuel and oxidizer velocity, etc.) similar to those employed in the work reported.

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

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.

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)

Magnesium carbide synthesis from methane and magnesium oxide - a potential methodology for natural gas conversion to premium fuels and chemicals

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

Diaz, A.F.; Modestino, A.J.; Howard, J.B.

1995-12-31

Diversification of the raw materials base for manufacturing premium fuels and chemicals offers U.S. and international consumers economic and strategic benefits. Extensive reserves of natural gas in the world provide a valuable source of clean gaseous fuel and chemical feedstock. Assuming the availability of suitable conversion processes, natural gas offers the prospect of improving flexibility in liquid fuels and chemicals manufacture, and thus, the opportunity to complement, supplement, or displace petroleum-based production as economic and strategic considerations require. The composition of natural gas varies from reservoir to reservoir but the principal hydrocarbon constituent is always methane (CH{sub 4}). With itsmore » high hydrogen-to-carbon ratio, methane has the potential to produce hydrogen or hydrogen-rich products. However, methane is a very chemically stable molecule and, thus, is not readily transformed to other molecules or easily reformed to its elements (H{sub 2} and carbon). In many cases, further research is needed to augment selectivity to desired product(s), increase single-pass conversions, or improve economics (e.g. there have been estimates of $50/bbl or more for liquid products) before the full potential of these methodologies can be realized on a commercial scale. With the trade-off between gas conversion and product selectivity, a major challenge common to many of these technologies is to simultaneously achieve high methane single-pass conversions and high selectivity to desired products. Based on the results of the scoping runs, there appears to be strong indications that a breakthrough has finally been achieved in that synthesis of magnesium carbides from MgO and methane in the arc discharge reactor has been demonstrated.« less

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.

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.

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT ...

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

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT FRAME AND SUPPORT PLATFORM, AND SAFETY MECHANISM ASSEMBLY (SPRING-LOADED HINGE). F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-1. INEL INDEX CODE NUMBER: 075 0701 60 851 151975. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... transuranic elements. Different technical processes can accomplish this separation. However, over the years Purex has become the most commonly used and accepted process. Purex involves the dissolution of... facilities have process functions similar to each other, including: irradiated fuel element chopping, fuel...

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

Analysis of physical-chemical processes governing SSME internal fluid flows

NASA Technical Reports Server (NTRS)

Singhal, A. K.; Owens, S. F.; Mukerjee, T.; Keeton, L. W.; Prakash, C.; Przekwas, A. J.

1984-01-01

The efforts to adapt CHAM's computational fluid dynamics code, PHOENICS, to the analysis of flow within the high pressure fuel turbopump (HPFTP) aft-platform seal cavity of the SSME are summarized. In particular, the special purpose PHOENICS satellite and ground station specifically formulated for this application are listed and described, and the preliminary results of the first part two-dimensional analyses are presented and discussed. Planned three-dimensional analyses are also briefly outlined. To further understand the mixing and combustion processes in the SSME fuelside preburners, a single oxygen-hydrogen jet element was investigated.

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.

Wavelength modulation spectroscopy near 5 μm for carbon monoxide sensing in a high-pressure kerosene-fueled liquid rocket combustor

NASA Astrophysics Data System (ADS)

Lee, Daniel D.; Bendana, Fabio A.; Schumaker, S. Alexander; Spearrin, R. Mitchell

2018-05-01

A laser absorption sensor was developed for carbon monoxide (CO) sensing in high-pressure, fuel-rich combustion gases associated with the internal conditions of hydrocarbon-fueled liquid bipropellant rockets. An absorption feature near 4.98 μm, comprised primarily of two rovibrational lines from the P-branch of the fundamental band, was selected to minimize temperature sensitivity and spectral interference with other combustion gas species at the extreme temperatures (> 3000 K) and pressures (> 50 atm) in the combustion chamber environment. A scanned wavelength modulation spectroscopy technique (1 f-normalized 2 f detection) is utilized to infer species concentration from CO absorption, and mitigate the influence of non-absorption transmission losses and noise associated with the harsh sooting combustor environment. To implement the sensing strategy, a continuous-wave distributed-feedback (DFB) quantum cascade laser (QCL) was coupled to a hollow-core optical fiber for remote mid-infrared light delivery to the test article, with high-bandwidth light detection by a direct-mounted photovoltaic detector. The method was demonstrated to measure time-resolved CO mole fraction over a range of oxidizer-to-fuel ratios and pressures (20-70 atm) in a single-element-injector RP-2-GOx rocket combustor.

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

NASA Technical Reports Server (NTRS)

Mcdonald, Gary H.

1988-01-01

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

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

Application of dual-fuel propulsion to a single stage AMLS vehicle

NASA Technical Reports Server (NTRS)

Lepsch, Roger A., Jr.; Stanley, Douglas O.; Unal, Resit

1993-01-01

As part of NASA's Advanced Manned Launch System (AMLS) study to determine a follow-on, or complement, to the Space Shuttle, a reusable single-stage-to-orbit concept utilizing dual-fuel rocket propulsion has been examined. Several dual-fuel propulsion concepts were investigated. These include: a separate engine concept combining Russian RD-170 kerosene-fueled engines with SSME-derivative engines; the kerosene and hydrogen-fueled Russian RD-701 engine concept; and a dual-fuel, dual-expander engine concept. Analysis to determine vehicle weight and size characteristics was performed using conceptual level design techniques. A response surface methodology for multidisciplinary design was utilized to optimize the dual-fuel vehicle concepts with respect to several important propulsion system and vehicle design parameters in order to achieve minimum empty weight. Comparisons were then made with a hydrogen-fueled reference, single-stage vehicle. The tools and methods employed in the analysis process are also summarized.

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.

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...

75 FR 71346 - Special Conditions: Boeing Model 787-8 Airplane; Lightning Protection of Fuel Tank Structure To...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-23

... Boeing Model 787-8 airplane will incorporate a fuel tank nitrogen generation system (NGS) that actively... ignition source in the fuel tank system could not result from any single failure, from any single failure... of fuel systems. We do not intend to apply the alternative standards used under these special...

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.

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).

Dissolver vessel bottom assembly

DOEpatents

Kilian, Douglas C.

1976-01-01

An improved bottom assembly is provided for a nuclear reactor fuel reprocessing dissolver vessel wherein fuel elements are dissolved as the initial step in recovering fissile material from spent fuel rods. A shock-absorbing crash plate with a convex upper surface is disposed at the bottom of the dissolver vessel so as to provide an annular space between the crash plate and the dissolver vessel wall. A sparging ring is disposed within the annular space to enable a fluid discharged from the sparging ring to agitate the solids which deposit on the bottom of the dissolver vessel and accumulate in the annular space. An inlet tangential to the annular space permits a fluid pumped into the annular space through the inlet to flush these solids from the dissolver vessel through tangential outlets oppositely facing the inlet. The sparging ring is protected against damage from the impact of fuel elements being charged to the dissolver vessel by making the crash plate of such a diameter that the width of the annular space between the crash plate and the vessel wall is less than the diameter of the fuel elements.

14 CFR 31.46 - Pressurized fuel systems.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Pressurized fuel systems. 31.46 Section 31... AIRWORTHINESS STANDARDS: MANNED FREE BALLOONS Design Construction § 31.46 Pressurized fuel systems. For pressurized fuel systems, each element and its connecting fittings and lines must be tested to an ultimate...

14 CFR 31.46 - Pressurized fuel systems.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Pressurized fuel systems. 31.46 Section 31... AIRWORTHINESS STANDARDS: MANNED FREE BALLOONS Design Construction § 31.46 Pressurized fuel systems. For pressurized fuel systems, each element and its connecting fittings and lines must be tested to an ultimate...

14 CFR 31.46 - Pressurized fuel systems.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Pressurized fuel systems. 31.46 Section 31... AIRWORTHINESS STANDARDS: MANNED FREE BALLOONS Design Construction § 31.46 Pressurized fuel systems. For pressurized fuel systems, each element and its connecting fittings and lines must be tested to an ultimate...

PLOT PLAN OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS ...

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

PLOT PLAN OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS AND PROPOSED LOCATION OF FUEL ELEMENT CUTTING FACILITY. INL DRAWING NUMBER 200-0603-00-706-051287. ALTERNATE ID NUMBER CPP-C-1287. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Thermionic fuel element for the S-prime reactor

NASA Astrophysics Data System (ADS)

Van Hagan, Thomas H.; Drees, Elizabeth A.

1993-01-01

Technical aspects of the thermionic fuel element (TFE) design proposed for the S-PRIME space nuclear power system are discussed. Topics covered include the rational for selecting a multicell TFE approach, a technical description of the S-PRIME TFE and its estimated performance, and the technology readiness of the design, which emphasizes techology maturity and low risk.

METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Layer, E.H. Jr.; Peet, C.S.

1962-01-23

A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

METHOD OF MANUFACTURE OF METAL ENCASED CORE MATERIAL

DOEpatents

Peters, E.J.

1963-06-01

A method of making reactor fuel elements in which the fissionable material is encased and sealed in steel or aluminum cladding having an enclosed channel from the fissionable material to the surface of the cladding at one end is described. Heat and pressure sufficient to bond the assembled fuel element are applied in a nonoxidizing atmosphere. (AEC)

FUEL ELEMENT AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.

1961-04-25

A nuclear fuel element in the form of a wire is reported. A bar of uranium is enclosed in a thin layer of aluminum and the composite is sheathed in beryllium, zirconium, or stainnless steel. The sheathed article is then drawn to wire form, heated to alloy the aluminum with both uranium and sheath, and finally cold worked.

10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

Code of Federal Regulations, 2010 CFR

2010-01-01

... performance and safety during reactor operation. Also, in all cases precise control of processes, procedures... elements include equipment that: (1) Normally comes in direct contact with, or directly processes or... pellets; (2) Automatic welding machines especially designed or prepared for welding end caps onto the fuel...

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

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

Talley, Darren G.

2017-04-01

This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code showsmore » good agreement between simulation and actual ACRR operations.« less

REACTOR

DOEpatents

Roman, W.G.

1961-06-27

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

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

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

Saqib, Naeem, E-mail: naeem.saqib@oru.se; Bäckström, Mattias, E-mail: mattias.backstrom@oru.se

Highlights: • Different solids waste incineration is discussed in grate fired and fluidized bed boilers. • We explained waste composition, temperature and chlorine effects on metal partitioning. • Excessive chlorine content can change oxide to chloride equilibrium partitioning the trace elements in fly ash. • Volatility increases with temperature due to increase in vapor pressure of metals and compounds. • In Fluidized bed boiler, most metals find themselves in fly ash, especially for wood incineration. - Abstract: Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of flymore » 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.« less

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Horning, W.A.; Lanning, D.D.; Donahue, D.J.

1959-10-01

A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

Aerosols emitted in underground mine air by diesel engine fueled with biodiesel.

PubMed

Bugarski, Aleksandar D; Cauda, Emanuele G; Janisko, Samuel J; Hummer, Jon A; Patts, Larry D

2010-02-01

Using biodiesel in place of petroleum diesel is considered by several underground metal and nonmetal mine operators to be a viable strategy for reducing the exposure of miners to diesel particulate matter. This study was conducted in an underground experimental mine to evaluate the effects of soy methyl ester biodiesel on the concentrations and size distributions of diesel aerosols and nitric oxides in mine air. The objective was to compare the effects of neat and blended biodiesel fuels with those of ultralow sulfur petroleum diesel. The evaluation was performed using a mechanically controlled, naturally aspirated diesel engine equipped with a muffler and a diesel oxidation catalyst. The effects of biodiesel fuels on size distributions and number and total aerosol mass concentrations were found to be strongly dependent on engine operating conditions. When fueled with biodiesel fuels, the engine contributed less to elemental carbon concentrations for all engine operating modes and exhaust configurations. The substantial increases in number concentrations and fraction of organic carbon (OC) in total carbon over the baseline were observed when the engine was fueled with biodiesel fuels and operated at light-load operating conditions. Size distributions for all test conditions were found to be single modal and strongly affected by engine operating conditions, fuel type, and exhaust configuration. The peak and total number concentrations as well as median diameter decreased with an increase in the fraction of biodiesel in the fuels, particularly for high-load operating conditions. The effects of the diesel oxidation catalyst, commonly deployed to counteract the potential increase in OC emissions due to use of biodiesel, were found to vary depending upon fuel formulation and engine operating conditions. The catalyst was relatively effective in reducing aerosol number and mass concentrations, particularly at light-load conditions, but also showed the potential for an increase in nitrogen dioxide concentrations at high-load modes.

An experimental study of fuel injection strategies in CAI gasoline engine

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

Hunicz, J.; Kordos, P.

2011-01-15

Combustion of gasoline in a direct injection controlled auto-ignition (CAI) single-cylinder research engine was studied. CAI operation was achieved with the use of the negative valve overlap (NVO) technique and internal exhaust gas re-circulation (EGR). Experiments were performed at single injection and split injection, where some amount of fuel was injected close to top dead centre (TDC) during NVO interval, and the second injection was applied with variable timing. Additionally, combustion at variable fuel-rail pressure was examined. Investigation showed that at fuel injection into recompressed exhaust fuel reforming took place. This process was identified via an analysis of the exhaust-fuelmore » mixture composition after NVO interval. It was found that at single fuel injection in NVO phase, its advance determined the heat release rate and auto-ignition timing, and had a strong influence on NO{sub X} emission. However, a delay of single injection to intake stroke resulted in deterioration of cycle-to-cycle variability. Application of split injection showed benefits of this strategy versus single injection. Examinations of different fuel mass split ratios and variable second injection timing resulted in further optimisation of mixture formation. At equal share of the fuel mass injected in the first injection during NVO and in the second injection at the beginning of compression, the lowest emission level and cyclic variability improvement were observed. (author)« less

PLUTONIUM FUEL RODS FOR PREPARATION OF TRANSPLUTONIC ELEMENTS

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

Bailey, W.J.

1962-02-01

Production by coextrusion of metallurgically bonded, Alclad, Al-7.35 wt% Pu alloy fuel rods with integral ends is discussed. The rods had a diameter of 0.94 in., length of, 60 in., and a nominal cladding thickness of 0.070 in. The Pu concentration was maintained at 83.3 g/rod. The coextrusion billets can be assembled with fuel cores in the as-cast condition. The casting hot-tops can be returned to the process stream. The process is useful for preparing transplutonic elements and production of high-exposure Pu. (J.R.D.)

FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Shaner, B.E.

1961-08-15

The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

Nuclear reactor fuel element with vanadium getter on cladding

DOEpatents

Johnson, Carl E.; Carroll, Kenneth G.

1977-01-01

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 vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel

NASA Astrophysics Data System (ADS)

Teague, Melissa; Tonks, Michael; Novascone, Stephen; Hayes, Steven

2014-01-01

Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON [1] fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez-Lucuta model was favorable.

Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

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

Melissa Teague; Michael Tonks; Stephen Novascone

2014-01-01

Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISONmore » fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez–Lucuta model was favorable.« less

Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

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

Sears, D.F.; Primeau, M.F.; Buchanan, C.

1997-08-01

Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinationsmore » showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.« less

Distribution and leaching characteristics of trace elements in ashes as a function of different waste fuels and incineration technologies.

PubMed

Saqib, Naeem; Bäckström, Mattias

2015-10-01

Impact of waste fuels (virgin/waste wood, mixed biofuel (peat, bark, wood chips) industrial, household, mixed waste fuel) and incineration technologies on partitioning and leaching behavior of trace elements has been investigated. Study included 4 grate fired and 9 fluidized boilers. Results showed that mixed waste incineration mostly caused increased transfer of trace elements to fly ash; particularly Pb/Zn. Waste wood incineration showed higher transfer of Cr, As and Zn to fly ash as compared to virgin wood. The possible reasons could be high input of trace element in waste fuel/change in volatilization behavior due to addition of certain waste fractions. The concentration of Cd and Zn increased in fly ash with incineration temperature. Total concentration in ashes decreased in order of Zn>Cu>Pb>Cr>Sb>As>Mo. The concentration levels of trace elements were mostly higher in fluidized boilers fly ashes as compared to grate boilers (especially for biofuel incineration). It might be attributed to high combustion efficiency due to pre-treatment of waste in fluidized boilers. Leaching results indicated that water soluble forms of elements in ashes were low with few exceptions. Concentration levels in ash and ash matrix properties (association of elements on ash particles) are crucial parameters affecting leaching. Leached amounts of Pb, Zn and Cr in >50% of fly ashes exceeded regulatory limit for disposal. 87% of chlorine in fly ashes washed out with water at the liquid to solid ratio 10 indicating excessive presence of alkali metal chlorides/alkaline earths. Copyright © 2015. Published by Elsevier B.V.

Evaluation of Uranium-235 Measurement Techniques

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

Kaspar, Tiffany C.; Lavender, Curt A.; Dibert, Mark W.

2017-05-23

Monolithic U-Mo fuel plates are rolled to final fuel element form from the original cast ingot, and thus any inhomogeneities in 235U distribution present in the cast ingot are maintained, and potentially exaggerated, in the final fuel foil. The tolerance for inhomogeneities in the 235U concentration in the final fuel element foil is very low. A near-real-time, nondestructive technique to evaluate the 235U distribution in the cast ingot is required in order to provide feedback to the casting process. Based on the technical analysis herein, gamma spectroscopy has been recommended to provide a near-real-time measure of the 235U distribution inmore » U-Mo cast plates.« less

Effect of Crystal Orientation on Fatigue Failure of Single Crystal Nickel Base Turbine Blade Superalloys

NASA Technical Reports Server (NTRS)

Arakere, Nagaraj K.; Swanson, Gregory R.

2000-01-01

High Cycle Fatigue (HCF) induced failures in aircraft gas-turbine engines is a pervasive problem affecting a wide range of components and materials. HCF is currently the primary cause of component failures in gas turbine aircraft engines. Turbine blades in high performance aircraft and rocket engines are increasingly being made of single crystal nickel superalloys. Single-crystal Nickel-base superalloys were developed to provide superior creep, stress rupture, melt resistance and thermomechanical fatigue capabilities over polycrystalline alloys previously used in the production of turbine blades and vanes. Currently the most widely used single crystal turbine blade superalloys are PWA 1480/1493 and PWA 1484. These alloys play an important role in commercial, military and space propulsion systems. PWA1493, identical to PWA1480, but with tighter chemical constituent control, is used in the NASA SSME (Space Shuttle Main Engine) alternate turbopump, a liquid hydrogen fueled rocket engine. Objectives for this paper are motivated by the need for developing failure criteria and fatigue life evaluation procedures for high temperature single crystal components, using available fatigue data and finite element modeling of turbine blades. Using the FE (finite element) stress analysis results and the fatigue life relations developed, the effect of variation of primary and secondary crystal orientations on life is determined, at critical blade locations. The most advantageous crystal orientation for a given blade design is determined. Results presented demonstrates that control of secondary and primary crystallographic orientation has the potential to optimize blade design by increasing its resistance to fatigue crack growth without adding additional weight or cost.

Single-Cylinder Diesel Engine Tests with Unstabilized Water-in-Fuel Emulsions

DOT National Transportation Integrated Search

1978-08-01

A single-cylinder, four-stroke cycle diesel engine was operated on unstabilized water-in-fuel emulsions. Two prototype devices were used to produce the emulsions on-line with the engine. More than 350 test points were run with baseline diesel fuel an...

Dynamic characteristics of single crystal SSME blades

NASA Technical Reports Server (NTRS)

Moss, L. A.; Smith, T. E.

1987-01-01

The Space Shuttle Main Engine (SSME) High Pressure Fuel Turbopump (HPFTP) blades are currently manufactured using a directionally solidified (DS) material, MAR-M-246+Hf. However, a necessity to reduce the occurrence of fatigue cracking within the DS blades has lead to an interest in the use of a single crystal (SC) material, PWA-1480. A study was initiated to determine the dynamic characteristics of the HPFTP blades made of SC material and find possible critical engine order excitations. This study examined both the first and second stage drive turbine blades of the HPFTP. The dynamic characterization was done analytically as well as experimentally. The analytical study examined the SC first stage HPFTP blade dynamic characteristics under typical operating conditions. The blades were analyzed using MSC/NASTRAN and a finite element model. Two operating conditions, 27500 RPM and 35000 RPM, were investigated.

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

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

Renfro, David; Chandler, David; Cook, David

2014-10-30

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 converted usingmore » 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

Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

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

Feldman, E.

When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodatemore » an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code.Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm{sup 2}. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow instability for these tests leads one to the conclusion that the same method should be able to predict the onset of flow instability in the MURR reasonably well.« less

Fuel injection and mixing systems having piezoelectric elements and methods of using the same

DOEpatents

Mao, Chien-Pei [Clive, IA; Short, John [Norwalk, IA; Klemm, Jim [Des Moines, IA; Abbott, Royce [Des Moines, IA; Overman, Nick [West Des Moines, IA; Pack, Spencer [Urbandale, IA; Winebrenner, Audra [Des Moines, IA

2011-12-13

A fuel injection and mixing system is provided that is suitable for use with various types of fuel reformers. Preferably, the system includes a piezoelectric injector for delivering atomized fuel, a gas swirler, such as a steam swirler and/or an air swirler, a mixing chamber and a flow mixing device. The system utilizes ultrasonic vibrations to achieve fuel atomization. The fuel injection and mixing system can be used with a variety of fuel reformers and fuel cells, such as SOFC fuel cells.

ADVANCED COURSE ON FUEL ELEMENTS FOR WATER COOLED POWER REACTORS, ORGANIZED BY THE NETHERLANDS'-NORWEGIAN REACTOR SCHOOL AT INSTITUTT FOR ATOMENERGI, KJELLER, NORWAY, 22nd AUGUST-3rd SEPTEMBER,1960. VOLUME III

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

Aas, S.; Barendregt, T.J.; Chesne, A.

1960-07-01

A series of lectures on fuel elements for water-cooled power reactors are presented. Topics covered include fabrication, properties, cladding, radiation damage, design, cycling, storage and transpont, and reprocessing. Separate records have been prepared for each section.

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND ...

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

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND LOWER FUEL ELEMENTS. TAKEN FROM TOP OF SHIELDING TANK WITH CAMERA POINTING TOWARDS BOTTOM OF TANK. SHOWS LADDER, SQUARE LIFTING FRAME, FUEL ELEMENT HOLDERS, AND CABLE CYLINDERS. INEL PHOTO NUMBER 65-5434, TAKEN OCTOBER 20, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Unsteady combustion of solid propellants

NASA Astrophysics Data System (ADS)

Chung, T. J.; Kim, P. K.

The oscillatory motions of all field variables (pressure, temperature, velocity, density, and fuel fractions) in the flame zone of solid propellant rocket motors are calculated using the finite element method. The Arrhenius law with a single step forward chemical reaction is used. Effects of radiative heat transfer, impressed arbitrary acoustic wave incidence, and idealized mean flow velocities are also investigated. Boundary conditions are derived at the solid-gas interfaces and at the flame edges which are implemented via Lagrange multipliers. Perturbation expansions of all governing conservation equations up to and including the second order are carried out so that nonlinear oscillations may be accommodated. All excited frequencies are calculated by means of eigenvalue analyses, and the combustion response functions corresponding to these frequencies are determined. It is shown that the use of isoparametric finite elements, Gaussian quadrature integration, and the Lagrange multiplier boundary matrix scheme offers a convenient approach to two-dimensional calculations.

Fuel and oxidizer valve assembly employs single solenoid actuator

NASA Technical Reports Server (NTRS)

1966-01-01

Valve assembly simultaneously starts or stops the flow of oxidizer and fuel from separate inlet channels to reaction control motors. The assembly combines an oxidizer shutoff valve and a fuel shutoff valve which are mechanically linked and operated by a single high-speed solenoid actuator.

Applying Thermodynamics to Fossil Fuels: Heats of Combustion from Elemental Compositions.

ERIC Educational Resources Information Center

Lloyd, William G.; Davenport, Derek A.

1980-01-01

Discussed are the calculations of heats of combustions of some selected fossil fuel compounds such as some foreign shale oils and United States coals. Heating values for coal- and petroleum-derived fuel oils are also presented. (HM)

Dissolution of Used Nuclear Fuel Using a TBP/N-Paraffin Solvent

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

Rudisill, T. S.; Shehee, T. C.; Jones, D. H.

2017-10-02

The dissolution of unirradiated used nuclear fuel (UNF) pellets pretreated for tritium removal was demonstrated using a tributly phosphate (TBP) solvent. Dissolution of pretreated fuel in TBP could potentially combine dissolution with two cycle of solvent extraction required for separating the actinides and lanthanides from other fission products. Dissolutions were performed using UNF surrogates prepared from both uranyl nitrate and uranium trioxide produced from the pretreatment process by adding selected actinide and stable fission product elements. In laboratory-scale experiments, the U dissolution efficiency ranged from 80-99+% for both the nitrate and oxide surrogate fuels. On average, 80% of the Pumore » and 50% of the Np and Am in the nitrate surrogate dissolved; however, little of the transuranic elements dissolved in the oxide form. The majority of the 3+ lanthanide elements dissolved. Only small amounts of Sr (0-1.6%) and Mo (0.1-1.7%) and essentially no Cs, Ru, Zr, or Pd dissolved.« less

An assessment of the benefits of the use of NASA developed fuel conservative technology in the US commercial aircraft fleet

NASA Technical Reports Server (NTRS)

1975-01-01

Cost and benefits of a fuel conservative aircraft technology program proposed by NASA are estimated. NASA defined six separate technology elements for the proposed program: (a) engine component improvement (b) composite structures (c) turboprops (d) laminar flow control (e) fuel conservative engine and (f) fuel conservative transport. There were two levels postulated: The baseline program was estimated to cost $490 million over 10 years with peak funding in 1980. The level two program was estimated to cost an additional $180 million also over 10 years. Discussions with NASA and with representatives of the major commercial airframe manufacturers were held to estimate the combinations of the technology elements most likely to be implemented, the potential fuel savings from each combination, and reasonable dates for incorporation of these new aircraft into the fleet.

Single chamber microbial fuel cell with Ni-Co cathode

NASA Astrophysics Data System (ADS)

Włodarczyk, Barbara; Włodarczyk, Paweł P.; Kalinichenko, Antonina

2017-10-01

The possibility of wastewater treatment and the parallel energy production using the Ni-Co alloy as cathode catalyst for single chamber microbial fuel cells is presented in this research. The research included a preparation of catalyst and comparison of COD, NH4+ and NO3- reduction in the reactor without aeration, with aeration and with using a single chamber microbial fuel cell with Ni-Co cathode. The reduction time for COD with the use of microbial fuel cell with the Ni-Co catalyst is similar to the reduction time with aeration. The current density (2.4 A·m-2) and amount of energy (0.48 Wh) obtained in MFC is low, but the obtained amount of energy allows elimination of the energy needed for reactor aeration. It has been shown that the Ni-Co can be used as cathode catalyst in single chamber microbial fuel cells.

Automated assembling of single fuel cell units for use in a fuel cell stack

NASA Astrophysics Data System (ADS)

Jalba, C. K.; Muminovic, A.; Barz, C.; Nasui, V.

2017-05-01

The manufacturing of PEMFC stacks (POLYMER ELEKTROLYT MEMBRAN Fuel Cell) is nowadays still done by hand. Over hundreds of identical single components have to be placed accurate together for the construction of a fuel cell stack. Beside logistic problems, higher total costs and disadvantages in weight the high number of components produce a higher statistic interference because of faulty erection or material defects and summation of manufacturing tolerances. The saving of costs is about 20 - 25 %. Furthermore, the total weight of the fuel cells will be reduced because of a new sealing technology. Overall a one minute cycle time has to be aimed per cell at the manufacturing of these single components. The change of the existing sealing concept to a bonded sealing is one of the important requisites to get an automated manufacturing of single cell units. One of the important steps for an automated gluing process is the checking of the glue application by using of an image processing system. After bonding the single fuel cell the sealing and electrical function can be checked, so that only functional and high qualitative cells can get into further manufacturing processes.

Method for measuring recovery of catalytic elements from fuel cells

DOEpatents

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley, NJ

2011-03-08

A method is provided for measuring the concentration of a catalytic clement in a fuel cell powder. The method includes depositing on a porous substrate at least one layer of a powder mixture comprising the fuel cell powder and an internal standard material, ablating a sample of the powder mixture using a laser, and vaporizing the sample using an inductively coupled plasma. A normalized concentration of catalytic element in the sample is determined by quantifying the intensity of a first signal correlated to the amount of catalytic element in the sample, quantifying the intensity of a second signal correlated to the amount of internal standard material in the sample, and using a ratio of the first signal intensity to the second signal intensity to cancel out the effects of sample size.

Dual-Fuel Propulsion in Single-Stage Advanced Manned Launch System Vehicle

NASA Technical Reports Server (NTRS)

Lepsch, Roger A., Jr.; Stanley, Douglas O.; Unal, Resit

1995-01-01

As part of the United States Advanced Manned Launch System study to determine a follow-on, or complement, to the Space Shuttle, a reusable single-stage-to-orbit concept utilizing dual-fuel rocket propulsion has been examined. Several dual-fuel propulsion concepts were investigated. These include: a separate-engine concept combining Russian RD-170 kerosene-fueled engines with space shuttle main engine-derivative engines: the kerosene- and hydrogen-fueled Russian RD-701 engine; and a dual-fuel, dual-expander engine. Analysis to determine vehicle weight and size characteristics was performed using conceptual-level design techniques. A response-surface methodology for multidisciplinary design was utilized to optimize the dual-fuel vehicles with respect to several important propulsion-system and vehicle design parameters, in order to achieve minimum empty weight. The tools and methods employed in the analysis process are also summarized. In comparison with a reference hydrogen- fueled single-stage vehicle, results showed that the dual-fuel vehicles were from 10 to 30% lower in empty weight for the same payload capability, with the dual-expander engine types showing the greatest potential.

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

2007-01-01

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

DOEpatents

Tomczuk, Zygmunt; Miller, William E.; Wolson, Raymond D.; Gay, Eddie C.

1991-01-01

An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

Temperature Profile in Fuel and Tie-Tubes for Nuclear Thermal Propulsion Systems

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

Vishal Patel

A finite element method to calculate temperature profiles in heterogeneous geometries of tie-tube moderated LEU nuclear thermal propulsion systems and HEU designs with tie-tubes is developed and implemented in MATLAB. This new method is compared to previous methods to demonstrate shortcomings in those methods. Typical methods to analyze peak fuel centerline temperature in hexagonal geometries rely on spatial homogenization to derive an analytical expression. These methods are not applicable to cores with tie-tube elements because conduction to tie-tubes cannot be accurately modeled with the homogenized models. The fuel centerline temperature directly impacts safety and performance so it must be predictedmore » carefully. The temperature profile in tie-tubes is also important when high temperatures are expected in the fuel because conduction to the tie-tubes may cause melting in tie-tubes, which may set maximum allowable performance. Estimations of maximum tie-tube temperature can be found from equivalent tube methods, however this method tends to be approximate and overly conservative. A finite element model of heat conduction on a unit cell can model spatial dependence and non-linear conductivity for fuel and tie-tube systems allowing for higher design fidelity of Nuclear Thermal Propulsion.« less

Continuous process electrorefiner

DOEpatents

Herceg, Joseph E [Naperville, IL; Saiveau, James G [Hickory Hills, IL; Krajtl, Lubomir [Woodridge, IL

2006-08-29

A new device is provided for the electrorefining of uranium in spent metallic nuclear fuels by the separation of unreacted zirconium, noble metal fission products, transuranic elements, and uranium from spent fuel rods. The process comprises an electrorefiner cell. The cell includes a drum-shaped cathode horizontally immersed about half-way into an electrolyte salt bath. A conveyor belt comprising segmented perforated metal plates transports spent fuel into the salt bath. The anode comprises the conveyor belt, the containment vessel, and the spent fuel. Uranium and transuranic elements such as plutonium (Pu) are oxidized at the anode, and, subsequently, the uranium is reduced to uranium metal at the cathode. A mechanical cutter above the surface of the salt bath removes the deposited uranium metal from the cathode.

Laser diagnostics for NTP fuel corrosion studies

NASA Technical Reports Server (NTRS)

Wantuck, Paul J.; Butt, D. P.; Sappey, A. D.

1993-01-01

Viewgraphs and explanations on laser diagnostics for nuclear thermal propulsion (NTP) fuel corrosion studies are presented. Topics covered include: NTP fuels; U-Zr-C system corrosion products; planar laser-induced fluorescence (PLIF); utilization of PLIF for corrosion product characterization of nuclear thermal rocket fuel elements under test; ZrC emission spectrum; and PLIF imaging of ZrC plume.

Method For Processing Spent (Trn,Zr)N Fuel

DOEpatents

Miller, William E.; Richmann, Michael K.

2004-07-27

A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

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

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

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

NASA Technical Reports Server (NTRS)

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

2013-01-01

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

Effects of FAME biodiesel and HVORD on emissions from an older-technology diesel engine.

PubMed

Bugarski, A D; Hummer, J A; Vanderslice, S E

2017-12-01

The results of laboratory evaluations were used to compare the potential of two alternative, biomass-derived fuels as a control strategy to reduce the exposure of underground miners to aerosols and gases emitted by diesel-powered equipment. The effects of fatty acid methyl ester (FAME) biodiesel and hydrotreated vegetable oil renewable diesel (HVORD) on criteria aerosol and gaseous emissions from an older-technology, naturally aspirated, mechanically controlled engine equipped with a diesel oxidation catalytic converter were compared with those of widely used petroleum-derived, ultralow-sulfur diesels (ULSDs). The emissions were characterized for four selected steady-state conditions. When fueled with FAME biodiesel and HVORD, the engine emitted less aerosols by total particulate mass, total carbon mass, elemental carbon mass and total number than when it was fueled with ULSDs. Compared with ULSDs, FAME biodiesel and HVORD produced aerosols that were characterized by single modal distributions, smaller count median diameters, and lower total and peak concentrations. For the majority of test cases, FAME biodiesel and HVORD favorably affected nitric oxide (NO) and adversely affected nitrogen dioxide (NO 2 ) generation. Therefore, the use of these alternative fuels appears to be a viable tool for the underground mining industry to address the issues related to emissions from diesel engines, and to transition toward more universal solutions provided by advanced engines with integrated exhaust after treatment technologies.

Effects of FAME biodiesel and HVORD on emissions from an older-technology diesel engine

PubMed Central

Bugarski, A.D.; Hummer, J.A.; Vanderslice, S.E.

2017-01-01

The results of laboratory evaluations were used to compare the potential of two alternative, biomass-derived fuels as a control strategy to reduce the exposure of underground miners to aerosols and gases emitted by diesel-powered equipment. The effects of fatty acid methyl ester (FAME) biodiesel and hydrotreated vegetable oil renewable diesel (HVORD) on criteria aerosol and gaseous emissions from an older-technology, naturally aspirated, mechanically controlled engine equipped with a diesel oxidation catalytic converter were compared with those of widely used petroleum-derived, ultralow-sulfur diesels (ULSDs). The emissions were characterized for four selected steady-state conditions. When fueled with FAME biodiesel and HVORD, the engine emitted less aerosols by total particulate mass, total carbon mass, elemental carbon mass and total number than when it was fueled with ULSDs. Compared with ULSDs, FAME biodiesel and HVORD produced aerosols that were characterized by single modal distributions, smaller count median diameters, and lower total and peak concentrations. For the majority of test cases, FAME biodiesel and HVORD favorably affected nitric oxide (NO) and adversely affected nitrogen dioxide (NO2) generation. Therefore, the use of these alternative fuels appears to be a viable tool for the underground mining industry to address the issues related to emissions from diesel engines, and to transition toward more universal solutions provided by advanced engines with integrated exhaust after treatment technologies. PMID:29348698

Whole-genome sequencing of the efficient industrial fuel-ethanol fermentative Saccharomyces cerevisiae strain CAT-1.

PubMed

Babrzadeh, Farbod; Jalili, Roxana; Wang, Chunlin; Shokralla, Shadi; Pierce, Sarah; Robinson-Mosher, Avi; Nyren, Pål; Shafer, Robert W; Basso, Luiz C; de Amorim, Henrique V; de Oliveira, Antonio J; Davis, Ronald W; Ronaghi, Mostafa; Gharizadeh, Baback; Stambuk, Boris U

2012-06-01

The Saccharomyces cerevisiae strains widely used for industrial fuel-ethanol production have been developed by selection, but their underlying beneficial genetic polymorphisms remain unknown. Here, we report the draft whole-genome sequence of the S. cerevisiae strain CAT-1, which is a dominant fuel-ethanol fermentative strain from the sugarcane industry in Brazil. Our results indicate that strain CAT-1 is a highly heterozygous diploid yeast strain, and the ~12-Mb genome of CAT-1, when compared with the reference S228c genome, contains ~36,000 homozygous and ~30,000 heterozygous single nucleotide polymorphisms, exhibiting an uneven distribution among chromosomes due to large genomic regions of loss of heterozygosity (LOH). In total, 58 % of the 6,652 predicted protein-coding genes of the CAT-1 genome constitute different alleles when compared with the genes present in the reference S288c genome. The CAT-1 genome contains a reduced number of transposable elements, as well as several gene deletions and duplications, especially at telomeric regions, some correlated with several of the physiological characteristics of this industrial fuel-ethanol strain. Phylogenetic analyses revealed that some genes were likely associated with traits important for bioethanol production. Identifying and characterizing the allelic variations controlling traits relevant to industrial fermentation should provide the basis for a forward genetics approach for developing better fermenting yeast strains.

Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

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

Primm, Trent; Ruggles, Arthur; Freels, James D

2009-03-01

A finite element computational approach to simulation of the High Flux Isotope Reactor (HFIR) Core Thermal-Fluid behavior is developed. These models were developed to facilitate design of a low enriched core for the HFIR, which will have different axial and radial flux profiles from the current HEU core and thus will require fuel and poison load optimization. This report outlines a stepwise implementation of this modeling approach using the commercial finite element code, COMSOL, with initial assessment of fuel, poison and clad conduction modeling capability, followed by assessment of mating of the fuel conduction models to a one dimensional fluidmore » model typical of legacy simulation techniques for the HFIR core. The model is then extended to fully couple 2-dimensional conduction in the fuel to a 2-dimensional thermo-fluid model of the coolant for a HFIR core cooling sub-channel with additional assessment of simulation outcomes. Finally, 3-dimensional simulations of a fuel plate and cooling channel are presented.« less

Evaluation of factors that affect diesel exhaust toxicity. Final report

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

Norbeck, J.M.; Smith, M.R.; Arey, J.

1998-07-01

The scope of this project was to obtain a preliminary assessment of the potential impact of the fuel formulation on the speciation and toxic components of diesel exhaust. The test bed was a Cummins L10 engine operating over the heavy-duty transient test cycle using three diesel fuels: a pre-1993 diesel fuel, a low aromatic diesel fuel, and an alternative formulation diesel fuel. The sampling/analysis plan included: determination of the criteria pollutant emission rates (THC, CO, NOx, and PM); determination of PM(10) and PM(2.5) emission rates; collection and analysis of particulate samples for elemental, inorganic ion and elemental/organic carbon analyses; collectionmore » of bas samples for VOC speciation analyses; collection of 2,4-dinitrophenylhydrazine (DNPH) cartridges for determination of oxygenates; collection of nitrosomorpholine with Thermosorb N cartridges; collection of semi-volatiles on PF/XAD and particulate samples for PAH, nitro-PAH, and mutagenicity studies; and collection and analysis of dioxins for the pre-1993 and alternative formulation diesel fuels.« less

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

NASA Technical Reports Server (NTRS)

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

2016-01-01

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

Performance comparison of low-temperature direct alcohol fuel cells with different anode catalysts

NASA Astrophysics Data System (ADS)

Zhou, W. J.; Zhou, B.; Li, W. Z.; Zhou, Z. H.; Song, S. Q.; Sun, G. Q.; Xin, Q.; Douvartzides, S.; Goula, M.; Tsiakaras, P.

Low-temperature polymer electrolyte membrane fuel cells directly fed by methanol and ethanol were investigated employing carbon supported Pt, PtSn and PtRu as anode catalysts, respectively. Employing Pt/C as anode catalyst, both direct methanol fuel cell (DMFC) and direct ethanol fuel cell (DEFC) showed poor performances even in presence of high Pt loading on anode. It was found that the addition of Ru or Sn to the Pt dramatically enhances the electro-oxidation of both methanol and ethanol. It was also found that the single cell adopting PtRu/C as anode shows better DMFC performance, while PtSn/C catalyst shows better DEFC performance. The single fuel cell using PtSn/C as anode catalyst at 90 °C shows similar power densities whenever fueled by methanol or ethanol. The cyclic voltammetry (CV) and single fuel cell tests indicated that PtRu is more suitable for DMFC while PtSn is more suitable for DEFC.

High performance platinum single atom electrocatalyst for oxygen reduction reaction

NASA Astrophysics Data System (ADS)

Liu, Jing; Jiao, Menggai; Lu, Lanlu; Barkholtz, Heather M.; Li, Yuping; Wang, Ying; Jiang, Luhua; Wu, Zhijian; Liu, Di-Jia; Zhuang, Lin; Ma, Chao; Zeng, Jie; Zhang, Bingsen; Su, Dangsheng; Song, Ping; Xing, Wei; Xu, Weilin; Wang, Ying; Jiang, Zheng; Sun, Gongquan

2017-07-01

For the large-scale sustainable implementation of polymer electrolyte membrane fuel cells in vehicles, high-performance electrocatalysts with low platinum consumption are desirable for use as cathode material during the oxygen reduction reaction in fuel cells. Here we report a carbon black-supported cost-effective, efficient and durable platinum single-atom electrocatalyst with carbon monoxide/methanol tolerance for the cathodic oxygen reduction reaction. The acidic single-cell with such a catalyst as cathode delivers high performance, with power density up to 680 mW cm-2 at 80 °C with a low platinum loading of 0.09 mgPt cm-2, corresponding to a platinum utilization of 0.13 gPt kW-1 in the fuel cell. Good fuel cell durability is also observed. Theoretical calculations reveal that the main effective sites on such platinum single-atom electrocatalysts are single-pyridinic-nitrogen-atom-anchored single-platinum-atom centres, which are tolerant to carbon monoxide/methanol, but highly active for the oxygen reduction reaction.

High performance platinum single atom electrocatalyst for oxygen reduction reaction

PubMed Central

Liu, Jing; Jiao, Menggai; Lu, Lanlu; Barkholtz, Heather M.; Li, Yuping; Wang, Ying; Jiang, Luhua; Wu, Zhijian; Liu, Di-jia; Zhuang, Lin; Ma, Chao; Zeng, Jie; Zhang, Bingsen; Su, Dangsheng; Song, Ping; Xing, Wei; Xu, Weilin; Wang, Ying; Jiang, Zheng; Sun, Gongquan

2017-01-01

For the large-scale sustainable implementation of polymer electrolyte membrane fuel cells in vehicles, high-performance electrocatalysts with low platinum consumption are desirable for use as cathode material during the oxygen reduction reaction in fuel cells. Here we report a carbon black-supported cost-effective, efficient and durable platinum single-atom electrocatalyst with carbon monoxide/methanol tolerance for the cathodic oxygen reduction reaction. The acidic single-cell with such a catalyst as cathode delivers high performance, with power density up to 680 mW cm−2 at 80 °C with a low platinum loading of 0.09 mgPt cm−2, corresponding to a platinum utilization of 0.13 gPt kW−1 in the fuel cell. Good fuel cell durability is also observed. Theoretical calculations reveal that the main effective sites on such platinum single-atom electrocatalysts are single-pyridinic-nitrogen-atom-anchored single-platinum-atom centres, which are tolerant to carbon monoxide/methanol, but highly active for the oxygen reduction reaction. PMID:28737170

Fully Burdened Cost of Fuel Using Input-Output Analysis

DTIC Science & Technology

2011-12-01

Distribution Model could be used to replace the current seven-step Fully Burdened Cost of Fuel process with a single step, allowing for less complex and...wide extension of the Bulk Fuels Distribution Model could be used to replace the current seven-step Fully Burdened Cost of Fuel process with a single...ABBREVIATIONS AEM Atlantic, Europe, and the Mediterranean AOAs Analysis of Alternatives DAG Defense Acquisition Guidebook DAU Defense Acquisition University

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

NASA Astrophysics Data System (ADS)

Govers, K.; Verwerft, M.

2016-09-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

Analysis and fabrication of tungsten CERMET materials for ultra-high temperature reactor applications via pulsed electric current sintering

NASA Astrophysics Data System (ADS)

Webb, Jonathan A.

The optimized development path for the fabrication of ultra-high temperature W-UO2 CERMET fuel elements were explored within this dissertation. A robust literature search was conducted, which concluded that a W-UO 2 fuel element must contain a fine tungsten microstructure and spherical UO2 kernels throughout the entire consolidation process. Combined Monte Carlo and Computational Fluid Dynamics (CFD) analysis were used to determine the effects of rhenium and gadolinia additions on the performance of W-UO 2 fuel elements at refractory temperatures and in dry and water submerged environments. The computational analysis also led to the design of quasi-optimized fuel elements that can meet thermal-hydraulic and neutronic requirements A rigorous set of experiments were conducted to determine if Pulsed Electric Current Sintering (PECS) can fabricate tungsten and W-Ce02 specimens to the required geometries, densities and microstructures required for high temperature fuel elements as well as determine the mechanisms involved within the PECS consolidation process. The CeO2 acts as a surrogate for UO 2 fuel kernels in these experiments. The experiments seemed to confirm that PECS consolidation takes place via diffusional mass transfer methods; however, the densification process is rapidly accelerated due to the effects of current densities within the consolidating specimen. Fortunately the grain growth proceeds at a traditional rate and the PECS process can yield near fully dense W and W-Ce02 specimens with a finer microstructure than other sintering techniques. PECS consolidation techniques were also shown to be capable of producing W-UO2 segments at near-prototypic geometries; however, great care must be taken to coat the fuel particles with tungsten prior to sintering. Also, great care must be taken to ensure that the particles remain spherical in geometry under the influence of a uniaxial stress as applied during PECS, which involves mixing different fuel kernel sizes in order to reduce the porosity in the initial green compact. Particle mixing techniques were also shown to be capable of producing consolidated CERMETs, but with a less than desirable microstructure. The work presented herin will help in the development of very high temperature reactors for terrestrial and space missions in the future.

Heterogeneous reduction of carbon dioxide by hydride-terminated silicon nanocrystals

PubMed Central

Sun, Wei; Qian, Chenxi; He, Le; Ghuman, Kulbir Kaur; Wong, Annabelle P. Y.; Jia, Jia; Jelle, Abdinoor A.; O'Brien, Paul G.; Reyes, Laura M.; Wood, Thomas E.; Helmy, Amr S.; Mims, Charles A.; Singh, Chandra Veer; Ozin, Geoffrey A.

2016-01-01

Silicon constitutes 28% of the earth's mass. Its high abundance, lack of toxicity and low cost coupled with its electrical and optical properties, make silicon unique among the semiconductors for converting sunlight into electricity. In the quest for semiconductors that can make chemicals and fuels from sunlight and carbon dioxide, unfortunately the best performers are invariably made from rare and expensive elements. Here we report the observation that hydride-terminated silicon nanocrystals with average diameter 3.5 nm, denoted ncSi:H, can function as a single component heterogeneous reducing agent for converting gaseous carbon dioxide selectively to carbon monoxide, at a rate of hundreds of μmol h−1 g−1. The large surface area, broadband visible to near infrared light harvesting and reducing power of SiH surface sites of ncSi:H, together play key roles in this conversion. Making use of the reducing power of nanostructured hydrides towards gaseous carbon dioxide is a conceptually distinct and commercially interesting strategy for making fuels directly from sunlight. PMID:27550234

Fast measurement of proton exchange membrane fuel cell impedance based on pseudo-random binary sequence perturbation signals and continuous wavelet transform

NASA Astrophysics Data System (ADS)

Debenjak, Andrej; Boškoski, Pavle; Musizza, Bojan; Petrovčič, Janko; Juričić, Đani

2014-05-01

This paper proposes an approach to the estimation of PEM fuel cell impedance by utilizing pseudo-random binary sequence as a perturbation signal and continuous wavelet transform with Morlet mother wavelet. With the approach, the impedance characteristic in the frequency band from 0.1 Hz to 500 Hz is identified in 60 seconds, approximately five times faster compared to the conventional single-sine approach. The proposed approach was experimentally evaluated on a single PEM fuel cell of a larger fuel cell stack. The quality of the results remains at the same level compared to the single-sine approach.

Effect of Crystal Orientation on Analysis of Single-Crystal, Nickel-Based Turbine Blade Superalloys

NASA Technical Reports Server (NTRS)

Swanson, G. R.; Arakere, N. K.

2000-01-01

High-cycle fatigue-induced failures in turbine and turbopump blades is a pervasive problem. Single-crystal nickel turbine blades are used because of their superior creep, stress rupture, melt resistance, and thermomechanical fatigue capabilities. Single-crystal materials have highly orthotropic properties making the position of the crystal lattice relative to the part geometry a significant and complicating factor. A fatigue failure criterion based on the maximum shear stress amplitude on the 24 octahedral and 6 cube slip systems is presented for single-crystal nickel superalloys (FCC crystal). This criterion greatly reduces the scatter in uniaxial fatigue data for PWA 1493 at 1,200 F in air. Additionally, single-crystal turbine blades used in the Space Shuttle main engine high pressure fuel turbopump/alternate turbopump are modeled using a three-dimensional finite element (FE) model. This model accounts for material orthotrophy and crystal orientation. Fatigue life of the blade tip is computed using FE stress results and the failure criterion that was developed. Stress analysis results in the blade attachment region are also presented. Results demonstrate that control of crystallographic orientation has the potential to significantly increase a component's resistance to fatigue crack growth without adding additional weight or cost.

Apparatus and method for removing particulate deposits from high temperature filters

DOEpatents

Nakaishi, Curtis V.; Holcombe, Norman T.; Micheli, Paul L.

1992-01-01

A combustion of a fuel-air mixture is used to provide a high-temperature and high-pressure pulse of gaseous combustion products for the back-flush cleaning of ceramic filter elements contained in a barrier filter system and utilized to separate particulates from particulate-laden process gases at high temperature and high pressure. The volume of gaseous combustion products provided by the combustion of the fuel-air mixture is preferably divided into a plurality of streams each passing through a sonic orifice and conveyed to the open end of each filter element as a high pressure pulse which passes through the filter elements and dislodges dust cake supported on a surface of the filter element.

Trace elements by instrumental neutron activation analysis for pollution monitoring

NASA Technical Reports Server (NTRS)

Sheibley, D. W.

1975-01-01

Methods and technology were developed to analyze 1000 samples/yr of coal and other pollution-related samples. The complete trace element analysis of 20-24 samples/wk averaged 3-3.5 man-hours/sample. The computerized data reduction scheme could identify and report data on as many as 56 elements. In addition to coal, samples of fly ash, bottom ash, crude oil, fuel oil, residual oil, gasoline, jet fuel, kerosene, filtered air particulates, ore, stack scrubber water, clam tissue, crab shells, river sediment and water, and corn were analyzed. Precision of the method was plus or minus 25% based on all elements reported in coal and other sample matrices. Overall accuracy was estimated at 50%.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William

2013-01-01

A key technology element in Nuclear Thermal Propulsion is the development of fuel materials and components which can withstand extremely high temperatures while being exposed to flowing hydrogen. NTREES provides a cost effective method for rapidly screening of candidate fuel components with regard to their viability for use in NTR systems. The NTREES is designed to mimic the conditions (minus the radiation) to which nuclear rocket fuel elements and other components would be subjected to during reactor operation. The NTREES consists of a water cooled ASME code stamped pressure vessel and its associated control hardware and instrumentation coupled with inductive heaters to simulate the heat provided by the fission process. The NTREES has been designed to safely allow hydrogen gas to be injected into internal flow passages of an inductively heated test article mounted in the chamber.

Fuel handling system for a nuclear reactor

DOEpatents

Saiveau, James G.; Kann, William J.; Burelbach, James P.

1986-01-01

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

Fuel handling system for a nuclear reactor

DOEpatents

Saiveau, James G.; Kann, William J.; Burelbach, James P.

1986-12-02

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

Nuclear fuel element nut retainer cup. [PWR

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

Walton, L.A.

1977-07-19

A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups aremore » partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant.« less

Physics Features of TRU-Fueled VHTRs

DOE PAGES

Lewis, Tom G.; Tsvetkov, Pavel V.

2009-01-01

The current waste management strategy for spent nuclear fuel (SNF) mandated by the US Congress is the disposal of high-level waste (HLW) in a geological repository at Yucca Mountain. Ongoing efforts on closed-fuel cycle options and difficulties in opening and safeguarding such a repository have led to investigations of alternative waste management strategies. One potential strategy for the US fuel cycle would be to make use of fuel loadings containing high concentrations of transuranic (TRU) nuclides in the next-generation reactors. The use of such fuels would not only increase fuel supply but could also potentially facilitate prolonged operation modes (viamore » fertile additives) on a single fuel loading. The idea is to approach autonomous operation on a single fuel loading that would allow marketing power units as nuclear batteries for worldwide deployment. Studies have already shown that high-temperature gas-cooled reactors (HTGRs) and their Generation IV (GEN IV) extensions, very-high-temperature reactors (VHTRs), have encouraging performance characteristics. This paper is focused on possible physics features of TRU-fueled VHTRs. One of the objectives of a 3-year U.S. DOE NERI project was to show that TRU-fueled VHTRs have the possibility of prolonged operation on a single fuel loading. A 3D temperature distribution was developed based on conceivable operation conditions of the 600 MWth VHTR design. Results of extensive criticality and depletion calculations with varying fuel loadings showed that VHTRs are capable for autonomous operation and HLW waste reduction when loaded with TRU fuel.« less

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

Effects of idle reduction technologies on real world fuel use and exhaust emissions of idling long-haul trucks.

PubMed

Frey, H Christopher; Kuo, Po-Yao; Villa, Charles

2009-09-01

Idling long-haul freight tucks may consume nearly one billion gallons of diesel fuel per year in the U.S. There is a need for real-world data by which to quantify avoided fuel use and emissions attributable to idle reduction techniques of auxiliary power units (APUs) and shore-power (SP). Field data were obtained from 20 APU-equipped and SP-compatible trucks observed during 2.8 million miles of travel in 42 states. Base engine fuel use and emission rates varied depending on ambient temperature. APU and SP energy use and emission rates varied depending on electrical load. APUs reduced idling fuel use and CO2 emissions for single and team drivers by 22 and 5% annually, respectively. SP offers greater reductions in energy use of 48% for single drivers, as well as in emissions, except for SO2. APUs were cost-effective for single drivers with a large number of APU usage hours per year, but not for team drivers or for single drivers with low APU utilization rates. The findings support more accurate assessments of avoided fuel use and emissions, and recommendations to encourage greater APU utilization by single drivers and to further develop infrastructure for SP.

Vented nuclear fuel element

DOEpatents

Grossman, Leonard N.; Kaznoff, Alexis I.

1979-01-01

A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

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

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

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

I. Glagolenko; D. Wachs; N. Woolstenhulme

2010-10-01

Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily duemore » to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.« less

Evaluation of Instrumentation for Measuring Undissolved Water in Aviation Turbine Fuels per ASTM D3240

DTIC Science & Technology

2015-11-05

Undissolved Water in Aviation Turbine Fuels per ASTM D3240 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Joel Schmitigal... water ) in Aviation Turbine Fuels per ASTM D3240 15. SUBJECT TERMS fuel, JP-8, aviation fuel, contamination, free water , undissolved water , Aqua-Glo 16...Michigan 48397-5000 Evaluation of Instrumentation for Measuring Undissolved Water in Aviation Turbine Fuels per ASTM D3240 Joel Schmitigal Force

APPARATUS FOR LOADING AND UNLOADING A MACHINE

DOEpatents

Payne, J.H. Jr.

1962-07-17

An arrangement for loading and unloading a nuclear reactor is described. Depleted fuel elements are removed from the reactor through one of a small number of holes in a shielding plug that is rotatably mounted in an eccentric annular plug rotatably mounted in the top of the reactor. The fuel elements removed are stored in a plurality of openings in a rotatable magazine or storage means rotatably mounted over the plugs. (AEC)

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Handwerk, J.H.; BAch, R.A.

1959-08-18

A method is described for preparing a reactor fuel element by forming a mixture of thorium dioxide and an oxide of uranium, the uranium being present. In an oxidation state at least as high as it is in U/sub 3/O/sub 8/, into a desired shape and firing in air at a temperature siifficiently high to reduce the higher uranium oxide to uranium dioxide.

Apparatus for testing high pressure injector elements

NASA Technical Reports Server (NTRS)

Myers, William Neill (Inventor); Scott, Ewell M. (Inventor); Forbes, John C. (Inventor); Shadoan, Michael D. (Inventor)

1995-01-01

An apparatus for testing and evaluating the spray pattern of high pressure fuel injector elements for use in supplying fuel to combustion engines is presented. Prior art fuel injector elements were normally tested by use of low pressure apparatuses which did not provide a purge to prevent mist from obscuring the injector element or to prevent frosting of the view windows; could utilize only one fluid during each test; and had their viewing ports positioned one hundred eighty (180 deg) apart, thus preventing optimum use of laser diagnostics. The high pressure fluid injector test apparatus includes an upper hub, an upper weldment or housing, a first clamp and stud/nut assembly for securing the upper hub to the upper weldment, a standoff assembly within the upper weldment, a pair of window housings having view glasses within the upper weldment, an injector block assembly and purge plate within the upper weldment for holding an injector element to be tested and evaluated, a lower weldment or housing, a second clamp and stud/nut assembly for securing the lower weldment to the upper hub, a third clamp and stud/nut assembly for securing the lower hub to the lower weldment, mechanisms for introducing fluid under high pressure for testing an injector element, and mechanisms for purging the apparatus to prevent frosting of view glasses within the window housings and to permit unobstructed viewing of the injector element.

Apparatus for testing high pressure injector elements

NASA Technical Reports Server (NTRS)

Myers, William Neill (Inventor); Scott, Ewell M. (Inventor); Forbes, John C. (Inventor); Shadoan, Michael D. (Inventor)

1993-01-01

An apparatus for testing and evaluating the spray pattern of high pressure fuel injector elements for use in supplying fuel to combustion engines is presented. Prior art fuel injector elements were normally tested by use of low pressure apparatuses which did not provide a purge to prevent mist from obscuring the injector element or to prevent frosting of the view windows; could utilize only one fluid during each test; and had their viewing ports positioned one hundred eighty (180 deg) apart, thus preventing optimum use of laser diagnostics. The high pressure fluid injector test apparatus includes an upper hub, an upper weldment or housing, a first clamp and stud/nut assembly for securing the upper hub to the upper weldment, a standoff assembly within the upper weldment, a pair of window housings having view glasses within the upper weldment, an injector block assembly and purge plate within the upper weldment for holding an injector element to be tested and evaluated, a lower weldment or housing, a second clamp and stud/nut assembly for securing the lower weldment to the upper weldment, a lower hub, a third clamp and stud/nut assembly for securing the lower hub to the lower weldment, mechanisms for introducing fluid under high pressure for testing an injector element, and mechanisms for purging the apparatus to prevent frosting of view glasses within the window housings and to permit unobstructed viewing of the injector element.

Fueling of magnetically confined plasmas by single- and two-stage repeating pneumatic pellet injectors

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Combs, S. K.; Foust, C. R.; Milora, S. L.

Advanced plasma fueling systems for magnetic fusion confinement experiments are under development at Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets to speeds in the kilometer-per-second range using single shot and repetitive pneumatic (light-gas gun) pellet injectors. The millimeter-to-centimeter size pellets enter the plasma and continuously ablate because of the plasma electron heat flux, depositing fuel atoms along the pellet trajectory. This fueling method allows direct fueling in the interior of the hot plasma and is more efficient than the alternative method of injecting room temperature fuel gas at the wall of the plasma vacuum chamber. Single-stage pneumatic injectors based on the light-gas gun concept have provided hydrogenic fuel pellets in the speed range of 1 to 2 km/s in single-shot injector designs. Repetition rates up to 5 Hz have been demonstrated in repetitive injector designs. Future fusion reactor-scale devices may need higher pellet velocities because of the larger plasma size and higher plasma temperatures. Repetitive two-stage pneumatic injectors are under development at ORNL to provide long-pulse plasma fueling in the 3 to 5 km/s speed range. Recently, a repeating, two-stage light-gas gun achieved repetitive operation at 1 Hz with speeds in the range of 2 to 3 km/s.

Solution combustion synthesis of strontium aluminate, SrAl2O4, powders: single-fuel versus fuel-mixture approach.

PubMed

Ianoş, Robert; Istratie, Roxana; Păcurariu, Cornelia; Lazău, Radu

2016-01-14

The solution combustion synthesis of strontium aluminate, SrAl2O4, via the classic single-fuel approach and the modern fuel-mixture approach was investigated in relation to the synthesis conditions, powder properties and thermodynamic aspects. The single-fuel approach (urea or glycine) did not yield SrAl2O4 directly from the combustion reaction. The absence of SrAl2O4 was explained by the low amount of energy released during the combustion process, in spite of the highly negative values of the standard enthalpy of reaction and standard Gibbs free energy. In the case of single-fuel recipes, the maximum combustion temperatures measured by thermal imaging (482 °C - urea, 941 °C - glycine) were much lower than the calculated adiabatic temperatures (1864 °C - urea, 2147 °C - glycine). The fuel-mixture approach (urea and glycine) clearly represented a better option, since (α,β)-SrAl2O4 resulted directly from the combustion reaction. The maximum combustion temperature measured in the case of a urea and glycine fuel mixture was the highest one (1559 °C), which was relatively close to the calculated adiabatic temperature (1930 °C). The addition of a small amount of flux, such as H3BO3, enabled the formation of pure α-SrAl2O4 directly from the combustion reaction.

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... uranium or enriching uranium in the isotope 235, zirconium tubes, heavy water or deuterium, nuclear-grade..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

Film bonded fuel cell interface configuration

DOEpatents

Kaufman, Arthur; Terry, Peter L.

1989-01-01

The present invention relates to improved elements for use in fuel cell stacks, and more particularly, to a stack having a corrosion-resistant, electrally conductive, fluid-impervious interface member therein.

Dehydrogenation of ammonia-borane by cationic Pd(II) and Ni(II) complexes in a nitromethane medium: hydrogen release and spent fuel characterization.

PubMed

Kim, Sung-Kwan; Hong, Sung-Ahn; Son, Ho-Jin; Han, Won-Sik; Michalak, Artur; Hwang, Son-Jong; Kang, Sang Ook

2015-04-28

A highly electrophilic cationic Pd(II) complex, [Pd(MeCN)4][BF4]2 (1), brings about the preferential activation of the B-H bond in ammonia-borane (NH3·BH3, AB). At room temperature, the reaction between 1 in CH3NO2 and AB in tetraglyme leads to Pd nanoparticles and formation of spent fuels of the general formula MeNHxBOy as reaction byproducts, while 2 equiv. of H2 is efficiently released per AB equiv. at room temperature within 60 seconds. For a mechanistic understanding of dehydrogenation by 1, the chemical structures of spent fuels were intensely characterized by a series of analyses such as elemental analysis (EA), X-ray photoelectron spectroscopy (XPS), solid state magic-angle-spinning (MAS) NMR spectra ((2)H, (13)C, (15)N, and (11)B), and cross polarization (CP) MAS methods. During AB dehydrogenation, the involvement of MeNO2 in the spent fuels showed that the mechanism of dehydrogenation catalyzed by 1 is different from that found in the previously reported results. This AB dehydrogenation derived from MeNO2 is supported by a subsequent digestion experiment of the AB spent fuel: B(OMe)3 and N-methylhydroxylamine ([Me(OH)N]2CH2), which are formed by the methanolysis of the AB spent fuel (MeNHxBOy), were identified by means of (11)B NMR and single crystal structural analysis, respectively. A similar catalytic behavior was also observed in the AB dehydrogenation catalyzed by a nickel catalyst, [Ni(MeCN)6][BF4]2 (2).

Determination of the /sup 237/Np(n, 2n) reaction cross section in the core of the BN-350 reactor

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

Goncharov, R.K.; Zvonarev, A.V.; Ivanov, V.I.

1987-05-01

Analysis of irradiated fuel (regular fuel elements and special ampule samples) makes it possible to obtain data on the neutron-physical characteristics that do not lend themselves to measurement in experiments with a low neutron flux. The authors studied samples of spent fuel from regular fuel elements and samples of /sup 236/U and /sup 237/Np irradiated in special ampules. We give the results of measurements of the ratio (sigma/sub n,2n/ + sigma/sub ..gamma..,n/)sigma/sub c/ for /sup 237/Np in different zones of the BN-350 reactor and the results of calculations performed using the neptunium group constants calculated from the data of themore » files ENDFB IV and ENDFB V« less

NACA Research on Slurry Fuels

NASA Technical Reports Server (NTRS)

Pinns, M L; Olson, W T; Barnett, H C; Breitwieser, R

1958-01-01

An extensive program was conducted to investigate the use of concentrated slurries of boron and magnesium in liquid hydrocarbon as fuels for afterburners and ramjet engines. Analytical calculations indicated that magnesium fuel would give greater thrust and that boron fuel would give greater range than are obtainable from jet hydrocarbon fuel alone. It was hoped that the use of these solid elements in slurry form would permit the improvement to be obtained without requiring unconventional fuel systems or combustors. Small ramjet vehicles fueled with magnesium slurry were flown successfully, but the test flights indicated that further improvement of combustors and fuel systems was needed.

Fully ceramic nuclear fuel and related methods

DOEpatents

Venneri, Francesco; Katoh, Yutai; Snead, Lance Lewis

2016-03-29

Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

Carbide fuel pin and capsule design for irradiations at thermionic temperatures

NASA Technical Reports Server (NTRS)

Siegel, B. L.; Slaby, J. G.; Mattson, W. F.; Dilanni, D. C.

1973-01-01

The design of a capsule assembly to evaluate tungsten-emitter - carbide-fuel combinations for thermionic fuel elements is presented. An inpile fuel pin evaluation program concerned with clad temperture, neutron spectrum, carbide fuel composition, fuel geometry,fuel density, and clad thickness is discussed. The capsule design was a compromise involving considerations between heat transfer, instrumentation, materials compatibility, and test location. Heat-transfer calculations were instrumental in determining the method of support of the fuel pin to minimize axial temperature variations. The capsule design was easily fabricable and utilized existing state-of-the-art experience from previous programs.

NORTHROP REACTOR. REVISION NO. 1 TO THE FINAL SAFEGUARDS REPORT

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

Duncan, J.M.; Shimizu, B.; Romine, R.A.

1962-10-01

Additions and changes related to the original application concerning construction and operation of the Northrop Reactor are given. Answers to 13 questions relative to the Final Safeguards Report are included. Answers are also included concerning 12 questions regarding receipt, possession, and storage of fuel elements. Other information is included concerning personnel changes and changes from Al-clad fuel elements to improved stainless steelclad hydride elements. It is concluded that the improved Northrop Reactor does not present any undue hazard to the health and safety of the operating personnel or the public. (J.R.D.)

Four-electron deoxygenative reductive coupling of carbon monoxide at a single metal site

NASA Astrophysics Data System (ADS)

Buss, Joshua A.; Agapie, Theodor

2016-01-01

Carbon dioxide is the ultimate source of the fossil fuels that are both central to modern life and problematic: their use increases atmospheric levels of greenhouse gases, and their availability is geopolitically constrained. Using carbon dioxide as a feedstock to produce synthetic fuels might, in principle, alleviate these concerns. Although many homogeneous and heterogeneous catalysts convert carbon dioxide to carbon monoxide, further deoxygenative coupling of carbon monoxide to generate useful multicarbon products is challenging. Molybdenum and vanadium nitrogenases are capable of converting carbon monoxide into hydrocarbons under mild conditions, using discrete electron and proton sources. Electrocatalytic reduction of carbon monoxide on copper catalysts also uses a combination of electrons and protons, while the industrial Fischer-Tropsch process uses dihydrogen as a combined source of electrons and electrophiles for carbon monoxide coupling at high temperatures and pressures. However, these enzymatic and heterogeneous systems are difficult to probe mechanistically. Molecular catalysts have been studied extensively to investigate the elementary steps by which carbon monoxide is deoxygenated and coupled, but a single metal site that can efficiently induce the required scission of carbon-oxygen bonds and generate carbon-carbon bonds has not yet been documented. Here we describe a molybdenum compound, supported by a terphenyl-diphosphine ligand, that activates and cleaves the strong carbon-oxygen bond of carbon monoxide, enacts carbon-carbon coupling, and spontaneously dissociates the resulting fragment. This complex four-electron transformation is enabled by the terphenyl-diphosphine ligand, which acts as an electron reservoir and exhibits the coordinative flexibility needed to stabilize the different intermediates involved in the overall reaction sequence. We anticipate that these design elements might help in the development of efficient catalysts for converting carbon monoxide to chemical fuels, and should prove useful in the broader context of performing complex multi-electron transformations at a single metal site.

Study of Compton suppression for use in spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Bender, Sarah

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

Reactor-based management of used nuclear fuel: assessment of major options.

PubMed

Finck, Phillip J; Wigeland, Roald A; Hill, Robert N

2011-01-01

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms). Copyright © 2010 Health Physics Society

Water Chemistry Control System for Recovery of Damaged and Degraded Spent Fuel

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

Sindelar, R.; Fisher, D.; Thomas, J.

2011-02-18

The International Atomic Energy Agency (IAEA) and the government of Serbia have led the project cosponsored by the U.S, Russia, European Commission, and others to repackage and repatriate approximately 8000 spent fuel elements from the RA reactor fuel storage basins at the VIN?A Institute of Nuclear Sciences to Russia for reprocessing. The repackaging and transportation activities were implemented by a Russian consortium which includes the Sosny Company, Tekhsnabeksport (TENEX) and Mayak Production Association. High activity of the water of the fuel storage basin posed serious risk and challenges to the fuel removal from storage containers and repackaging for transportation. Themore » risk centered on personnel exposure, even above the basin water, due to the high water activity levels caused by Cs-137 leached from fuel elements with failed cladding. A team of engineers from the U.S. DOE-NNSA's Global Threat Reduction Initiative, the Vinca Institute, and the IAEA performed the design, development, and deployment of a compact underwater water chemistry control system (WCCS) to remove the Cs-137 from the basin water and enable personnel safety above the basin water for repackaging operations. Key elements of the WCCS system included filters, multiple columns containing an inorganic sorbent, submersible pumps and flow meters. All system components were designed to be remotely serviceable and replaceable. The system was assembled and successfully deployed at the Vinca basin to support the fuel removal and repackaging activities. Following the successful operations, the Cs-137 is now safely contained and consolidated on the zeolite sorbent used in the columns of the WCCS, and the fuel has been removed from the basins. This paper reviews the functional requirements, design, and deployment of the WCCS.« less

FUEL SUBASSEMBLY CONSTRUCTION FOR RADIAL FLOW IN A NUCLEAR REACTOR

DOEpatents

Treshow, M.

1962-12-25

An assembly of fuel elements for a boiling water reactor arranged for radial flow of the coolant is described. The ingress for the coolant is through a central header tube, perforated with parallel circumferertial rows of openings each having a lip to direct the coolant flow downward. Around the central tube there are a number of equally spaced concentric trays, closely fitiing the central header tube. Cylindrical fuel elements are placed in a regular pattern around the central tube, piercing the trays. A larger tube encloses the arrangement, with space provided for upward flow of coolart beyond the edge of the trays. (AEC)

Comparison of particle sizes between 238PuO 2 before aqueous processing, after aqueous processing, and after ball milling

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

Mulford, Roberta Nancy

Particle sizes determined for a single lot of incoming Russian fuel and for a lot of fuel after aqueous processing are compared with particle sizes measured on fuel after ball-milling. The single samples of each type are believed to have particle size distributions typical of oxide from similar lots, as the processing of fuel lots is fairly uniform. Variation between lots is, as yet, uncharacterized. Sampling and particle size measurement methods are discussed elsewhere.

Issues and Potential Program on Denatured Fuel Utilization.

DTIC Science & Technology

1978-12-01

HTGR fuel develop - ment program ; 4. coated particles of (U,Th)02 have been extensively tested as potential HTGR fuels . A detailed summary of the...current scrap and waste treatment requirements. dBase case for all HTGR (Prismatic Fuel Element) cases based on data in "Summary Program Plan...Alternate Program for HTGR Fuel Recycle," April 11, 1975, Draft. 19 a --- AC8NCi09 The principal factors that result in a nominally-higher cost for

Seed and blanket fuel arrangement for dual-phase nuclear reactors

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

Congdon, S.P.; Fawcett, R.M.

1992-09-22

This patent describes a fuel management method for a dual-phase nuclear reactor, it comprises: installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture, each of the bundles having a predetermined group of fuel elements; operating the reactor a first time; shutting down the reactor; reinstalling the fuel bundle at a second core location accessed by coolant through a relatively large aperture; and operating the reactor a second time.

Development of burnup dependent fuel rod model in COBRA-TF

NASA Astrophysics Data System (ADS)

Yilmaz, Mine Ozdemir

The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.

METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

DOEpatents

Smith, R.R.; Echo, M.W.; Doe, C.B.

1963-12-31

A process and a device for the continuous monitoring of fuel elements while in use in a liquid-metal-cooled, argonblanketed nuclear reactor are presented. A fraction of the argon gas is withdrawn, contacted with a negative electrical charge for attraction of any alkali metal formed from argon by neutron reaction, and recycled into the reactor. The electrical charge is introduced into water, and the water is examined for radioactive alkali metals. (AEC)

SUPERHEATING IN A BOILING WATER REACTOR

DOEpatents

Treshow, M.

1960-05-31

A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

DOEpatents

Rasor, N.S.; Hirsch, R.L.

1963-12-01

The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

DOEpatents

Roake, W.E.

1958-08-19

A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

Thermo-Physics Technical Note No. 60: thermal analysis of SNAP 10A reactor core during atmospheric reentry and resulting core disintegration and fuel element separation

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

Mouradian, E.M.

1966-02-16

A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

METHOD OF MAKING WIRE FUEL ELEMENTS

DOEpatents

Zambrow, J.L.

1960-08-01

A method is given for making a nuclear reactor fuel element in the form of a uranium-bearing wire clad with zirconium. A uranium bar is enclosed in a zirconium sheath which is coated with an oxide of magnesium, beryllium, or zirconium. The sheathed bar is then placed in a steel tube and reduced to the desired diameter by swaging at 800 to 900 deg C, after which the steel and oxide are removed.

CALANDRIA TYPE SODIUM GRAPHITE REACTOR

DOEpatents

Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

1964-02-11

A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

Catalytic thermal barrier coatings

DOEpatents

Kulkarni, Anand A.; Campbell, Christian X.; Subramanian, Ramesh

2009-06-02

A catalyst element (30) for high temperature applications such as a gas turbine engine. The catalyst element includes a metal substrate such as a tube (32) having a layer of ceramic thermal barrier coating material (34) disposed on the substrate for thermally insulating the metal substrate from a high temperature fuel/air mixture. The ceramic thermal barrier coating material is formed of a crystal structure populated with base elements but with selected sites of the crystal structure being populated by substitute ions selected to allow the ceramic thermal barrier coating material to catalytically react the fuel-air mixture at a higher rate than would the base compound without the ionic substitutions. Precious metal crystallites may be disposed within the crystal structure to allow the ceramic thermal barrier coating material to catalytically react the fuel-air mixture at a lower light-off temperature than would the ceramic thermal barrier coating material without the precious metal crystallites.

Method of preparing gas tags for identification of single and multiple failures of nuclear reactor fuel assemblies

DOEpatents

McCormick, Norman J.

1976-01-01

For use in the identification of failed fuel assemblies in a nuclear reactor, the ratios of the tag gas isotopic concentrations are located on curved surfaces to enable the ratios corresponding to failure of a single fuel assembly to be distinguished from those formed from any combination of two or more failed assemblies.

Diesel surrogate fuels for engine testing and chemical-kinetic modeling: Compositions and properties

DOE PAGES

Mueller, Charles J.; Cannella, William J.; Bays, J. Timothy; ...

2016-01-07

The primary objectives of this work were to formulate, blend, and characterize a set of four ultralow-sulfur diesel surrogate fuels in quantities sufficient to enable their study in single-cylinder-engine and combustion-vessel experiments. The surrogate fuels feature increasing levels of compositional accuracy (i.e., increasing exactness in matching hydrocarbon structural characteristics) relative to the single target diesel fuel upon which the surrogate fuels are based. This approach was taken to assist in determining the minimum level of surrogate-fuel compositional accuracy that is required to adequately emulate the performance characteristics of the target fuel under different combustion modes. For each of the fourmore » surrogate fuels, an approximately 30 L batch was blended, and a number of the physical and chemical properties were measured. In conclusion, this work documents the surrogate-fuel creation process and the results of the property measurements.« less

Diesel Surrogate Fuels for Engine Testing and Chemical-Kinetic Modeling: Compositions and Properties

PubMed Central

Mueller, Charles J.; Cannella, William J.; Bays, J. Timothy; Bruno, Thomas J.; DeFabio, Kathy; Dettman, Heather D.; Gieleciak, Rafal M.; Huber, Marcia L.; Kweon, Chol-Bum; McConnell, Steven S.; Pitz, William J.; Ratcliff, Matthew A.

2016-01-01

The primary objectives of this work were to formulate, blend, and characterize a set of four ultralow-sulfur diesel surrogate fuels in quantities sufficient to enable their study in single-cylinder-engine and combustion-vessel experiments. The surrogate fuels feature increasing levels of compositional accuracy (i.e., increasing exactness in matching hydrocarbon structural characteristics) relative to the single target diesel fuel upon which the surrogate fuels are based. This approach was taken to assist in determining the minimum level of surrogate-fuel compositional accuracy that is required to adequately emulate the performance characteristics of the target fuel under different combustion modes. For each of the four surrogate fuels, an approximately 30 L batch was blended, and a number of the physical and chemical properties were measured. This work documents the surrogate-fuel creation process and the results of the property measurements. PMID:27330248

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

Mueller, Charles J.; Cannella, William J.; Bays, J. Timothy

The primary objectives of this work were to formulate, blend, and characterize a set of four ultralow-sulfur diesel surrogate fuels in quantities sufficient to enable their study in single-cylinder-engine and combustion-vessel experiments. The surrogate fuels feature increasing levels of compositional accuracy (i.e., increasing exactness in matching hydrocarbon structural characteristics) relative to the single target diesel fuel upon which the surrogate fuels are based. This approach was taken to assist in determining the minimum level of surrogate-fuel compositional accuracy that is required to adequately emulate the performance characteristics of the target fuel under different combustion modes. For each of the fourmore » surrogate fuels, an approximately 30 L batch was blended, and a number of the physical and chemical properties were measured. In conclusion, this work documents the surrogate-fuel creation process and the results of the property measurements.« less

Effect of Fuel Temperature Profile on Eigenvalue Calculations

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

Greifenkamp, Tom E; Clarno, Kevin T; Gehin, Jess C

2008-01-01

Use of an average fuel temperature is a current practice when modeling fuel for eigenvalue (k-inf) calculations. This is an approximation, as it is known from Heat-transfer methods that a fuel pin having linear power q', will have a temperature that varies radially and has a maximum temperature at the center line [1]. This paper describes an investigation into the effects on k-inf and isotopic concentrations of modeling a fuel pin using a single average temperature versus a radially varying fuel temperature profile. The axial variation is not discussed in this paper. A single fuel pin was modeled having 1,more » 3, 5, 8, or 10 regions of equal volumes (areas). Fig. 1 shows a model of a 10-ring fuel pin surrounded by a gap and then cladding.« less

High performance platinum single atom electrocatalyst for oxygen reduction reaction

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

Liu, Jing; Jiao, Menggai; Lu, Lanlu

For the large-scale sustainable implementation of polymer electrolyte membrane fuel cells in vehicles, high-performance electrocatalysts with low platinum consumption are desirable for use as cathode material during the oxygen reduction reaction in fuel cells. Here we report a carbon black-supported cost-effective, efficient and durable platinum single-atom electrocatalyst with carbon monoxide/methanol tolerance for the cathodic oxygen reduction reaction. The acidic single-cell with such a catalyst as cathode delivers high performance, with power density up to 680 mW cm –2 at 80 °C with a low platinum loading of 0.09 mgPt cm –2, corresponding to a platinum utilization of 0.13 gPt kWmore » –1 in the fuel cell. Good fuel cell durability is also observed. As a result, theoretical calculations reveal that the main effective sites on such platinum single-atom electrocatalysts are single-pyridinic-nitrogen-atom-anchored single-platinum-atom centres, which are tolerant to carbon monoxide/methanol, but highly active for the oxygen reduction reaction.« less

High performance platinum single atom electrocatalyst for oxygen reduction reaction

DOE PAGES

Liu, Jing; Jiao, Menggai; Lu, Lanlu; ...

2017-07-24

For the large-scale sustainable implementation of polymer electrolyte membrane fuel cells in vehicles, high-performance electrocatalysts with low platinum consumption are desirable for use as cathode material during the oxygen reduction reaction in fuel cells. Here we report a carbon black-supported cost-effective, efficient and durable platinum single-atom electrocatalyst with carbon monoxide/methanol tolerance for the cathodic oxygen reduction reaction. The acidic single-cell with such a catalyst as cathode delivers high performance, with power density up to 680 mW cm –2 at 80 °C with a low platinum loading of 0.09 mgPt cm –2, corresponding to a platinum utilization of 0.13 gPt kWmore » –1 in the fuel cell. Good fuel cell durability is also observed. As a result, theoretical calculations reveal that the main effective sites on such platinum single-atom electrocatalysts are single-pyridinic-nitrogen-atom-anchored single-platinum-atom centres, which are tolerant to carbon monoxide/methanol, but highly active for the oxygen reduction reaction.« less

Radiation Re-solution Calculation in Uranium-Silicide Fuels

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

Matthews, Christopher; Andersson, Anders David Ragnar; Unal, Cetin

The release of fission gas from nuclear fuels is of primary concern for safe operation of nuclear power plants. Although the production of fission gas atoms can be easily calculated from the fission rate in the fuel and the average yield of fission gas, the actual diffusion, behavior, and ultimate escape of fission gas from nuclear fuel depends on many other variables. As fission gas diffuses through the fuel grain, it tends to collect into intra-granular bubbles, as portrayed in Figure 1.1. These bubbles continue to grow due to absorption of single gas atoms. Simultaneously, passing fission fragments can causemore » collisions in the bubble that result in gas atoms being knocked back into the grain. This so called “re-solution” event results in a transient equilibrium of single gas atoms within the grain. As single gas atoms progress through the grain, they will eventually collect along grain boundaries, creating inter-granular bubbles. As the inter-granular bubbles grow over time, they will interconnect with other grain-face bubbles until a pathway is created to the outside of the fuel surface, at which point the highly pressurized inter-granular bubbles will expel their contents into the fuel plenum. This last process is the primary cause of fission gas release. From the simple description above, it is clear there are several parameters that ultimately affect fission gas release, including the diffusivity of single gas atoms, the absorption and knockout rate of single gas atoms in intra-granular bubbles, and the growth and interlinkage of intergranular bubbles. Of these, the knockout, or re-solution rate has an particularly important role in determining the transient concentration of single gas atoms in the grain. The re-solution rate will be explored in the following sections with regards to uranium-silicide fuels in order to support future models of fission gas bubble behavior.« less

A New Turbo-shaft Engine Control Law during Variable Rotor Speed Transient Process

NASA Astrophysics Data System (ADS)

Hua, Wei; Miao, Lizhen; Zhang, Haibo; Huang, Jinquan

2015-12-01

A closed-loop control law employing compressor guided vanes is firstly investigated to solve unacceptable fuel flow dynamic change in single fuel control for turbo-shaft engine here, especially for rotorcraft in variable rotor speed process. Based on an Augmented Linear Quadratic Regulator (ALQR) algorithm, a dual-input, single-output robust control scheme is proposed for a turbo-shaft engine, involving not only the closed loop adjustment of fuel flow but also that of compressor guided vanes. Furthermore, compared to single fuel control, some digital simulation cases using this new scheme about variable rotor speed have been implemented on the basis of an integrated system of helicopter and engine model. The results depict that the command tracking performance to the free turbine rotor speed can be asymptotically realized. Moreover, the fuel flow transient process has been significantly improved, and the fuel consumption has been dramatically cut down by more than 2% while keeping the helicopter level fight unchanged.

Two-stage earth-to-orbit vehicles with dual-fuel propulsion in the Orbiter

NASA Technical Reports Server (NTRS)

Martin, J. A.

1982-01-01

Earth-to-orbit vehicle studies of future replacements for the Space Shuttle are needed to guide technology development. Previous studies that have examined single-stage vehicles have shown advantages for dual-fuel propulsion. Previous two-stage system studies have assumed all-hydrogen fuel for the Orbiters. The present study examined dual-fuel Orbiters and found that the system dry mass could be reduced with this concept. The possibility of staging the booster at a staging velocity low enough to allow coast-back to the launch site is shown to be beneficial, particularly in combination with a dual-fuel Orbiter. An engine evaluation indicated the same ranking of engines as did a previous single-stage study. Propane and RP-1 fuels result in lower vehicle dry mass than methane, and staged-combustion engines are preferred over gas-generator engines. The sensitivity to the engine selection is less for two-stage systems than for single-stage systems.

Kinetic Studies of Plasma Chemical Fuel Oxidation in Nanosecond Pulsed Discharges by Single and Two Photon Laser Induced Fluorescence

DTIC Science & Technology

2013-07-01

31st ICPIG, July 14-19, 2013, Granada , Spain Kinetic Studies of Plasma Chemical Fuel Oxidation in Nanosecond Pulsed Discharges by Single and...31st) (ICPIG) Held in Granada , Spain on 14-19 July 2013 14. ABSTRACT Single and two photon Laser Induced Fluorescence (LIF) spectroscopy is used for...Standard Form 298 (Rev. 8-98) Prescribed by ANSI Std Z39-18 31st ICPIG, July 14-19, 2013, Granada , Spain preheat the fuel-air mixture to the furnace

The Conceptual Design for a Fuel Assembly of a New Research Reactor

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

Ryu, J-S.; Cho, Y-G.; Yoon, D-B.

2004-10-06

A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibrationmore » characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.« less

Sources of PM10 Air Pollution in Rural Area in the Vicinity of a Highway In Žilina Selfgoverning Region, Slovakia

NASA Astrophysics Data System (ADS)

Jandačka, Dušan

2015-05-01

Particulate matter results as an aftermath of numerous distinctive processes in the atmosphere and they become a part of everyday life. Their harmful effect and impact on the ambient environment is determined predominantly by the presence of various chemical substances and elements. The chemical composition of these particles (organic and elemental carbon, mineral dust, sea aerosols, secondary particles, especially sulphates and nitrates, heavy metals and further elements) is mainly impacted on by their origin, whereas the primary source of the particulate matter is determined and specified by the profile of chemical elements and substances. Particulate Matter (PM) may originate in various natural resources or anthropogenic sources. Among the natural sources sea salt is to be counted on, dust of the earth crust, pollen and volcanic ashes. Anthropogenic sources do include, predominantly, burning fossil fuels in the fossil-fuel power plants, local heating of households, burning liquefied fossil fuels in the combustion engines of vehicles, noncombustion related emissions as a result of vehicular traffic, resuspension of the road-traffic-related dust.

Acute Dermal Irritation Study of Six Jet Fuels in New Zealand White Rabbits: Comparison of Four Bio-Based Jet Fuels with Two Petroleum JP-8 Fuels

DTIC Science & Technology

2014-02-01

NA 5c. PROGRAM ELEMENT NUMBER 62202F 6. AUTHOR(S) Sterner, Teresa R.1; Hurley, Jonathon M.2; Edwards, James T.3; Shafer, Linda M.4; Mattie , David R... Mattie , D.R. 2014. Acute Dermal Irritation Study of Ten Jet Fuels in New Zealand White Rabbits: Comparison of Synthetic and Bio -Based Jet Fuels with...AFRL-RH-WP-TR-2014-0046 ACUTE DERMAL IRRITATION STUDY OF SIX JET FUELS IN NEW ZEALAND WHITE RABBITS: COMPARISON OF FOUR BIO -BASED JET FUELS

Automated fuel pin loading system

DOEpatents

Christiansen, David W.; Brown, William F.; Steffen, Jim M.

1985-01-01

An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inserted as a batch prior to welding of end caps by one of two disclosed welding systems.

Automated fuel pin loading system

DOEpatents

Christiansen, D.W.; Brown, W.F.; Steffen, J.M.

An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inerted as a batch prior to welding of end caps by one of two disclosed welding systems.

Film bonded fuel cell interface configuration

DOEpatents

Kaufman, Arthur; Terry, Peter L.

1985-01-01

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. A multi-layer arrangement for the interface provides bridging electrical contact with a hot-pressed resin filling the void space.

Process for making film-bonded fuel cell interfaces

DOEpatents

Kaufman, Arthur; Terry, Peter L.

1990-07-03

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. A multi-layer arrangement for the interface provides bridging electrical contact with a hot-pressed resin filling the void space.

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

Zee, Ralph; Schindler, Anton; Duke, Steve

The objective of this project is to conduct research to determine the feasibility of using alternate fuel sources for the production of cement. Successful completion of this project will also be beneficial to other commercial processes that are highly energy intensive. During this report period, we have completed all the subtasks in the preliminary survey. Literature searches focused on the types of alternative fuels currently used in the cement industry around the world. Information was obtained on the effects of particular alternative fuels on the clinker/cement product and on cement plant emissions. Federal regulations involving use of waste fuels weremore » examined. Information was also obtained about the trace elements likely to be found in alternative fuels, coal, and raw feeds, as well as the effects of various trace elements introduced into system at the feed or fuel stage on the kiln process, the clinker/cement product, and concrete made from the cement. The experimental part of this project involves the feasibility of a variety of alternative materials mainly commercial wastes to substitute for coal in an industrial cement kiln in Lafarge NA and validation of the experimental results with energy conversion consideration.« less

Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

2009-08-01

Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

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

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validatedmore » using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.« less

Blister Threshold Based Thermal Limits for the U-Mo Monolithic Fuel System

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

D. M. Wachs; I. Glagolenko; F. J. Rice

2012-10-01

Fuel failure is most commonly induced in research and test reactor fuel elements by exposure to an under-cooled or over-power condition that results in the fuel temperature exceeding a critical threshold above which blisters form on the plate. These conditions can be triggered by normal operational transients (i.e. temperature overshoots that may occur during reactor startup or power shifts) or mild upset events (e.g., pump coastdown, small blockages, mis-loading of fuel elements into higher-than-planned power positions, etc.). The rise in temperature has a number of general impacts on the state of a fuel plate that include, for example, stress relaxationmore » in the cladding (due to differential thermal expansion), softening of the cladding, increased mobility of fission gases, and increased fission-gas pressure in pores, all of which can encourage the formation of blisters on the fuel-plate surface. These blisters consist of raised regions on the surface of fuel plates that occur when the cladding plastically deforms in response to fission-gas pressure in large pores in the fuel meat and/or mechanical buckling of the cladding over damaged regions in the fuel meat. The blister temperature threshold decreases with irradiation because the mechanical properties of the fuel plate degrade while under irradiation (due to irradiation damage and fission-product accumulation) and because the fission-gas inventory progressively increases (and, thus, so does the gas pressure in pores).« less

Analysis of long-time operation of micro-cogeneration unit with fuel cell

NASA Astrophysics Data System (ADS)

Patsch, Marek; Čaja, Alexander

2015-05-01

Micro-cogeneration is cogeneration with small performance, with maximal electric power up to 50 kWe. On the present, there are available small micro-cogeneration units with small electric performance, about 1 kWe, which are usable also in single family houses or flats. These micro-cogeneration units operate on principle of conventional combustion engine, Stirling engine, steam engine or fuel cell. Micro-cogeneration units with fuel cells are new progressive developing type of units for single family houses. Fuel cell is electrochemical device which by oxidation-reduction reaction turn directly chemical energy of fuel to electric power, secondary products are pure water and thermal energy. The aim of paper is measuring and evaluation of operation parameters of micro-cogeneration unit with fuel cell which uses natural gas as a fuel.

Chemical composition and source of fine and nanoparticles from recent direct injection gasoline passenger cars: Effects of fuel and ambient temperature

NASA Astrophysics Data System (ADS)

Fushimi, Akihiro; Kondo, Yoshinori; Kobayashi, Shinji; Fujitani, Yuji; Saitoh, Katsumi; Takami, Akinori; Tanabe, Kiyoshi

2016-01-01

Particle number, mass, and chemical compositions (i.e., elemental carbon (EC), organic carbon (OC), elements, ions, and organic species) of fine particles emitted from four of the recent direct injection spark ignition (DISI) gasoline passenger cars and a port fuel injection (PFI) gasoline passenger car were measured under Japanese official transient mode (JC08 mode). Total carbon (TC = EC + OC) dominated the particulate mass (90% on average). EC dominated the TC for both hot and cold start conditions. The EC/TC ratios were 0.72 for PFI and 0.88-1.0 (average = 0.92) for DISI vehicles. A size-resolved chemical analysis of a DISI car revealed that the major organic components were the C20-C28 hydrocarbons for both the accumulation-mode particles and nanoparticles. Contribution of engine oil was estimated to be 10-30% for organics and the sum of the measured elements. The remaining major fraction likely originated from gasoline fuel. Therefore, it is suggested that soot (EC) also mainly originated from the gasoline. In experiments using four fuels at three ambient temperatures, the emission factors of particulate mass were consistently higher with regular gasoline than with premium gasoline. This result suggest that the high content of less-volatile compounds in fuel increase particulate emissions. These results suggest that focusing on reducing fuel-derived EC in the production process of new cars would effectively reduce particulate emission from DISI cars.

Study of advanced fuel system concepts for commercial aircraft and engines

NASA Technical Reports Server (NTRS)

Versaw, E. F.; Brewer, G. D.; Byers, W. D.; Fogg, H. W.; Hanks, D. E.; Chirivella, J.

1983-01-01

The impact on a commercial transport aircraft of using fuels which have relaxed property limits relative to current commercial jet fuel was assessed. The methodology of the study is outlined, fuel properties are discussed, and the effect of the relaxation of fuel properties analyzed. Advanced fuel system component designs that permit the satisfactory use of fuel with the candidate relaxed properties in the subject aircraft are described. The two fuel properties considered in detail are freezing point and thermal stability. Three candidate fuel system concepts were selected and evaluated in terms of performance, cost, weight, safety, and maintainability. A fuel system that incorporates insulation and electrical heating elements on fuel tank lower surfaces was found to be most cost effective for the long term.

Casting evaluation of U-Zr alloy system fuel slug for SFR prepared by injection casting method

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

Song, Hoon; Kim, Jong-Hwan; Kim, Ki-Hwan

2013-07-01

Metal fuel slugs of U-Pu-Zr alloys for Sodium-cooled Fast Reactor (SFR) have conventionally been fabricated by a vacuum injection casting method. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long-lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. In order tomore » prevent the evaporation of volatile elements such as Am, alternative fabrication methods of metal fuel slugs have been studied applying gravity casting, and improved injection casting in KAERI, including melting under inert atmosphere. And then, metal fuel slugs were examined with casting soundness, density, chemical analysis, particle size distribution and microstructural characteristics. Based on these results there is a high level of confidence that Am losses will also be effectively controlled by application of a modest amount of overpressure. A surrogate fuel slug was generally soundly cast by improved injection casting method, melted fuel material under inert atmosphere.« less

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

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

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

2013-07-01

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

Performance of a single layer fuel cell based on a mixed proton-electron conducting composite

NASA Astrophysics Data System (ADS)

Zagórski, Krzysztof; Wachowski, Sebastian; Szymczewska, Dagmara; Mielewczyk-Gryń, Aleksandra; Jasiński, Piotr; Gazda, Maria

2017-06-01

Many of the challenges in solid oxide fuel cell technology stem from chemical and mechanical incompatibilities between the anode, cathode and electrolyte materials. Numerous attempts have been made to identify compatible materials. Here, these challenges are circumvented by the introduction of a working single layer fuel cell, fabricated from a composite of proton conducting BaCe0.6Zr0.2Y0.2O3-δ and a mixture of semiconducting oxides - Li2O, NiO, and ZnO. Structural and electrical properties of the composite, related to its fuel cell performance are investigated. The single layer fuel cell shows a maximum OCV of 0.83 V and a peak power density of 3.86 mW cm-2 at 600 °C. Activation and mass transport losses are identified as the major limiting factor for efficiency and power output.