Apparatus and method for mixing fuel in a gas turbine nozzle

DOEpatents

Johnson, Thomas Edward; Ziminsky, Willy Steve; Berry, Jonathan Dwight

2014-08-12

A nozzle includes a fuel plenum and an air plenum downstream of the fuel plenum. A primary fuel channel includes an inlet in fluid communication with the fuel plenum and a primary air port in fluid communication with the air plenum. Secondary fuel channels radially outward of the primary fuel channel include a secondary fuel port in fluid communication with the fuel plenum. A shroud circumferentially surrounds the secondary fuel channels. A method for mixing fuel and air in a nozzle prior to combustion includes flowing fuel to a fuel plenum and flowing air to an air plenum downstream of the fuel plenum. The method further includes injecting fuel from the fuel plenum through a primary fuel passage, injecting fuel from the fuel plenum through secondary fuel passages, and injecting air from the air plenum through the primary fuel passage.

Performance of low smeared density sodium-cooled fast reactor metal fuel

DOE PAGES

Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; ...

2015-06-17

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

Compact plasma accelerator

NASA Technical Reports Server (NTRS)

Foster, John E. (Inventor)

2004-01-01

A compact plasma accelerator having components including a cathode electron source, an anodic ionizing gas source, and a magnetic field that is cusped. The components are held by an electrically insulating body having a central axis, a top axial end, and a bottom axial end. The cusped magnetic field is formed by a cylindrical magnet having an axis of rotation that is the same as the axis of rotation of the insulating body, and magnetized with opposite poles at its two axial ends; and an annular magnet coaxially surrounding the cylindrical magnet, magnetized with opposite poles at its two axial ends such that a top axial end has a magnetic polarity that is opposite to the magnetic polarity of a top axial end of the cylindrical magnet. The ionizing gas source is a tubular plenum that has been curved into a substantially annular shape, positioned above the top axial end of the annular magnet such that the plenum is centered in a ring-shaped cusp of the magnetic field generated by the magnets. The plenum has one or more capillary-like orifices spaced around its top such that an ionizing gas supplied through the plenum is sprayed through the one or more orifices. The plenum is electrically conductive and is positively charged relative to the cathode electron source such that the plenum functions as the anode; and the cathode is positioned above and radially outward relative to the plenum.

Cooling circuit for a gas turbine bucket and tip shroud

DOEpatents

Willett, Fred Thomas

2004-07-13

An open cooling circuit for a gas turbine airfoil and associated tip shroud includes a first group of cooling holes internal to the airfoil and extending in a radially outward direction generally along a leading edge of the airfoil; a second group of cooling holes internal to the airfoil and extending in a radially outward direction generally along a trailing edge of the airfoil. A common plenum is formed in the tip shroud in direct communication with the first and second group of cooling holes, but a second plenum may be provided for the second group of radial holes. A plurality of exhaust holes extends from the plenum(s), through the tip shroud and opening along a peripheral edge of the tip shroud.

Pressure suppression system

DOEpatents

Gluntz, D.M.

1994-10-04

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein. 3 figs.

Pressure suppression system

DOEpatents

Gluntz, Douglas M.

1994-01-01

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein.

Laser inertial fusion-based energy: Neutronic design aspects of a hybrid fusion-fission nuclear energy system

NASA Astrophysics Data System (ADS)

Kramer, Kevin James

This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 mum of tungsten to mitigate x-ray damage. The first wall is cooled by Li17Pb83 eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li17Pb 83 flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li17Pb83, separated from the Li17Pb83 by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF2), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles having a packing fraction of 20% in 2 cm diameter fuel pebbles. The fission blanket is cooled by the same radial flibe flow that travels through perforated ODS walls to the reflector blanket. This reflector blanket is 75 cm thick comprised of 2 cm diameter graphite pebbles cooled by flibe. The flibe extraction plenum surrounds the reflector bed. Detailed neutronics designs studies are performed to arrive at the described design. The LFFH engine thermal power is controlled using a technique of adjusting the 6Li/7Li enrichment in the primary and secondary coolants. The enrichment adjusts system thermal power in the design by increasing tritium production while reducing fission. To perform the simulations and design of the LFFH engine, a new software program named LFFH Nuclear Control (LNC) was developed in C++ to extend the functionality of existing neutron transport and depletion software programs. Neutron transport calculations are performed with MCNP5. Depletion calculations are performed using Monteburns 2.0, which utilizes ORIGEN 2.0 and MCNP5 to perform a burnup calculation. LNC supports many design parameters and is capable of performing a full 3D system simulation from initial startup to full burnup. It is able to iteratively search for coolant 6Li enrichments and resulting material compositions that meet user defined performance criteria. LNC is utilized throughout this study for time dependent simulation of the LFFH engine. Two additional methods were developed to improve the computation efficiency of LNC calculations. These methods, termed adaptive time stepping and adaptive mesh refinement were incorporated into a separate stand alone C++ library name the Adaptive Burnup Library (ABL). The ABL allows for other client codes to call and utilize its functionality. Adaptive time stepping is useful for automatically maximizing the size of the depletion time step while maintaining a desired level of accuracy. Adaptive meshing allows for analysis of fixed fuel configurations that would normally require a computationally burdensome number of depletion zones. Alternatively, Adaptive Mesh Refinement (AMR) adjusts the depletion zone size according to the variation in flux across the zone or fractional contribution to total absorption or fission. A parametric analysis on a fully mixed fuel core was performed using the LNC and ABL code suites. The resulting system parameters are found to optimize performance metrics using a 20 MT DU fuel load with a 20% TRISO packing and a 300 im kernel diameter operated with a fusion input power of 500 MW and a fission blanket gain of 4.0. LFFH potentially offers a proliferation resistant technology relative to other nuclear energy systems primarily because of no need for fuel enrichment or reprocessing. A figure of merit of the material attractiveness is examined and it is found that the fuel is effectively contaminated to an unattractive level shortly after the system is started due to fission product and minor actinide build up.

Monitoring system for a liquid-cooled nuclear fission reactor

DOEpatents

DeVolpi, Alexander

1987-01-01

A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

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.

Analysis of a Nuclear Enhanced Airbreathing Rocket for Earth to Orbit Applications

NASA Technical Reports Server (NTRS)

Adams, Robert B.; Landrum, D. Brian; Brown, Norman (Technical Monitor)

2001-01-01

The proposed engine concept is the Nuclear Enhanced Airbreathing Rocket (NEAR). The NEAR concept uses a fission reactor to thermally heat a propellant in a rocket plenum. The rocket is shrouded, thus the exhaust mixes with ingested air to provide additional thermal energy through combustion. The combusted flow is then expanded through a nozzle to provide thrust.

Radial midframe baffle for can-annular combustor arrangement having tangentially oriented combustor cans

DOEpatents

Rodriguez, Jose L.

2015-09-15

A can-annular gas turbine engine combustion arrangement (10), including: a combustor can (12) comprising a combustor inlet (38) and a combustor outlet circumferentially and axially offset from the combustor inlet; an outer casing (24) defining a plenum (22) in which the combustor can is disposed; and baffles (70) configured to divide the plenum into radial sectors (72) and configured to inhibit circumferential motion of compressed air (16) within the plenum.

Swirling midframe flow for gas turbine engine having advanced transitions

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

Montgomery, Matthew D.; Charron, Richard C.; Rodriguez, Jose L.

A gas turbine engine can-annular combustion arrangement (10), including: an axial compressor (82) operable to rotate in a rotation direction (60); a diffuser (100, 110) configured to receive compressed air (16) from the axial compressor; a plenum (22) configured to receive the compressed air from the diffuser; a plurality of combustor cans (12) each having a combustor inlet (38) in fluid communication with the plenum, wherein each combustor can is tangentially oriented so that a respective combustor inlet is circumferentially offset from a respective combustor outlet in a direction opposite the rotation direction; and an airflow guiding arrangement (80) configuredmore » to impart circumferential motion to the compressed air in the plenum in the direction opposite the rotation direction.« less

Active Aircraft Pylon Noise Control System

NASA Technical Reports Server (NTRS)

Thomas, Russell H. (Inventor); Czech, Michael J (Inventor); Elmiligui, Alaa A. (Inventor)

2015-01-01

An active pylon noise control system for an aircraft includes a pylon structure connecting an engine system with an airframe surface of the aircraft and having at least one aperture to supply a gas or fluid therethrough, an intake portion attached to the pylon structure to intake a gas or fluid, a regulator connected with the intake portion via a plurality of pipes, to regulate a pressure of the gas or fluid, a plenum chamber formed within the pylon structure and connected with the regulator, and configured to receive the gas or fluid as regulated by the regulator, and a plurality of injectors in communication with the plenum chamber to actively inject the gas or fluid through the plurality of apertures of the pylon structure.

Active Aircraft Pylon Noise Control System

NASA Technical Reports Server (NTRS)

Thomas, Russell H. (Inventor); Czech, Michael J. (Inventor); Elmiligui, Alaa A. (Inventor)

2017-01-01

An active pylon noise control system for an aircraft includes a pylon structure connecting an engine system with an airframe surface of the aircraft and having at least one aperture to supply a gas or fluid therethrough, an intake portion attached to the pylon structure to intake a gas or fluid, a regulator connected with the intake portion via a plurality of pipes, to regulate a pressure of the gas or fluid, a plenum chamber formed within the pylon structure and connected with the regulator, and configured to receive the gas or fluid as regulated by the regulator, and a plurality of injectors in communication with the plenum chamber to actively inject the gas or fluid through the plurality of apertures of the pylon structure.

Gas arc constriction for plasma arc welding

NASA Technical Reports Server (NTRS)

McGee, William F. (Inventor); Rybicki, Daniel J. (Inventor)

1994-01-01

A welding torch for plasma arc welding apparatus has an inert gas applied circumferentially about the arc column externally of the constricting nozzle so as to apply a constricting force on the arc after it has exited the nozzle orifice and downstream of the auxiliary shielding gas. The constricting inert gas is supplied to a plenum chamber about the body of the torch and exits through a series of circumferentially disposed orifices in an annular wall forming a closure at the forward end of the constricting gas plenum chamber. The constricting force of the circumferential gas flow about the arc concentrates and focuses the arc column into a more narrow and dense column of energy after exiting the nozzle orifice so that the arc better retains its energy density prior to contacting the workpiece.

CFD Analysis of Upper Plenum Flow for a Sodium-Cooled Small Modular Reactor

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

Kraus, A.; Hu, R.

2015-01-01

Upper plenum flow behavior is important for many operational and safety issues in sodium fast reactors. The Prototype Gen-IV Sodium Fast Reactor (PGSFR), a pool-type, 150 MWe output power design, was used as a reference case for a detailed characterization of upper plenum flow for normal operating conditions. Computational Fluid Dynamics (CFD) simulation was utilized with detailed geometric modeling of major structures. Core outlet conditions based on prior system-level calculations were mapped to approximate the outlet temperatures and flow rates for each core assembly. Core outlet flow was found to largely bypass the Upper Internal Structures (UIS). Flow curves overmore » the shield and circulates within the pool before exiting the plenum. Cross-flows and temperatures were evaluated near the core outlet, leading to a proposed height for the core outlet thermocouples to ensure accurate assembly-specific temperature readings. A passive scalar was used to evaluate fluid residence time from core outlet to IHX inlet, which can be used to assess the applicability of various methods for monitoring fuel failure. Additionally, the gas entrainment likelihood was assessed based on the CFD simulation results. Based on the evaluation of velocity gradients and turbulent kinetic energies and the available gas entrainment criteria in the literature, it was concluded that significant gas entrainment is unlikely for the current PGSFR design.« less

Solid fuel combustion system for gas turbine engine

DOEpatents

Wilkes, Colin; Mongia, Hukam C.

1993-01-01

A solid fuel, pressurized fluidized bed combustion system for a gas turbine engine includes a carbonizer outside of the engine for gasifying coal to a low Btu fuel gas in a first fraction of compressor discharge, a pressurized fluidized bed outside of the engine for combusting the char residue from the carbonizer in a second fraction of compressor discharge to produce low temperature vitiated air, and a fuel-rich, fuel-lean staged topping combustor inside the engine in a compressed air plenum thereof. Diversion of less than 100% of compressor discharge outside the engine minimizes the expense of fabricating and maintaining conduits for transferring high pressure and high temperature gas and incorporation of the topping combustor in the compressed air plenum of the engine minimizes the expense of modifying otherwise conventional gas turbine engines for solid fuel, pressurized fluidized bed combustion.

Apparatus for mixing fuel in a gas turbine nozzle

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

Barker, Carl Robert

A fuel nozzle in a combustion turbine engine that includes: a fuel plenum defined between an circumferentially extending shroud and axially by a forward tube-sheet and an aft tube-sheet; and a mixing-tube that extends across the fuel plenum that defines a passageway connecting an inlet formed through the forward tube-sheet and an outlet formed through the aft tube-sheet, the mixing-tube comprising one or more fuel ports that fluidly communicate with the fuel plenum. The mixing-tube may include grooves on an outer surface, and be attached to the forward tube-sheet by a connection having a fail-safe leakage path.

Enhancing the x-ray output of a single-wire explosion with a gas-puff based plasma opening switch

NASA Astrophysics Data System (ADS)

Engelbrecht, Joseph T.; Ouart, Nicholas D.; Qi, Niansheng; de Grouchy, Philip W.; Shelkovenko, Tatiana A.; Pikuz, Sergey A.; Banasek, Jacob T.; Potter, William M.; Rocco, Sophia V.; Hammer, David A.; Kusse, Bruce R.; Giuliani, John L.

2018-02-01

We present experiments performed on the 1 MA COBRA generator using a low density, annular, gas-puff z-pinch implosion as an opening switch to rapidly transfer a current pulse into a single metal wire on axis. This gas-puff on axial wire configuration was studied for its promise as an opening switch and as a means of enhancing the x-ray output of the wire. We demonstrate that current can be switched from the gas-puff plasma into the wire, and that the timing of the switch can be controlled by the gas-puff plenum backing pressure. X-ray detector measurements indicate that for low plenum pressure Kr or Xe shots with a copper wire, this configuration can offer a significant enhancement in the peak intensity and temporal distribution of radiation in the 1-10 keV range.

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could actually help extend the life of a fuel pin in a "long pin" reactor design to a higher peak fuel burnup.

Cooler and particulate separator for an off-gas stack

DOEpatents

Wright, George T.

1992-01-01

An off-gas stack for a melter comprising an air conduit leading to two sets of holes, one set injecting air into the off-gas stack near the melter plenum and the second set injecting air downstream of the first set. The first set injects air at a compound angle, having both downward and tangential components, to create a reverse vortex flow, counter to the direction of flow of gas through the stack and also along the periphery of the stack interior surface. Air from the first set of holes pervents recirculation zones from forming and the attendant accumulation of particulate deposits on the wall of the stack and will also return to the plenum any particulate swept up in the gas entering the stack. The second set of holes injects air in the same direction as the gas in the stack to compensate for the pressure drop and to prevent the concentration of condensate in the stack. A set of sprayers, receiving water from a second conduit, is located downstream of the second set of holes and sprays water into the gas to further cool it.

Fractal Model of Fission Product Release in Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stankunas, Gediminas

2012-09-01

A model of fission gas migration in nuclear fuel pellet is proposed. Diffusion process of fission gas in granular structure of nuclear fuel with presence of inter-granular bubbles in the fuel matrix is simulated by fractional diffusion model. The Grunwald-Letnikov derivative parameter characterizes the influence of porous fuel matrix on the diffusion process of fission gas. A finite-difference method for solving fractional diffusion equations is considered. Numerical solution of diffusion equation shows correlation of fission gas release and Grunwald-Letnikov derivative parameter. Calculated profile of fission gas concentration distribution is similar to that obtained in the experimental studies. Diffusion of fission gas is modeled for real RBMK-1500 fuel operation conditions. A functional dependence of Grunwald-Letnikov derivative parameter with fuel burn-up is established.

Combustion Dynamics Behavior in a Single-Element Lean Direct Injection (LDI) Gas Turbine Combustor

DTIC Science & Technology

2014-06-01

Constant mass inflow from a choked orifice Exit Boundary Condition Choked nozzle Diameter of combustor 50.8 mm Diameter of air plenum 25.4 mm A...schematic of the LDI combustor is shown in Fig. 1. It comprises an air inlet section, air plenum, swirler- venturi- injector assembly, combustion chamber...and exit nozzle . Air, heated with an 80 kW electrical heater, enters the combustor through a slotted choked orifice plate, designed to minimize

METHOD OF PRODUCING AND ACCELERATING AN ION BEAM

NASA Technical Reports Server (NTRS)

Foster, John E. (Inventor)

2005-01-01

A method of producing and accelerating an ion beam comprising the steps of providing a magnetic field with a cusp that opens in an outward direction along a centerline that passes through a vertex of the cusp: providing an ionizing gas that sprays outward through at least one capillary-like orifice in a plenum that is positioned such that the orifice is on the centerline in the cusp, outward of the vortex of the cusp; providing a cathode electron source, and positioning it outward of the orifice and off of the centerline; and positively charging the plenum relative to the cathode electron source such that the plenum functions as m anode. A hot filament may be used as the cathode electron source, and permanent magnets may be used to provide the magnetic field.

Cooler and particulate separator for an off-gas stack

DOEpatents

Wright, G.T.

1991-04-08

This report describes an off-gas stack for a melter, furnace or reaction vessel comprising an air conduit leading to two sets of holes, one set injecting air into the off-gas stack near the melter plenum and the second set injecting air downstream of the first set. The first set injects air at a compound angle, having both downward and tangential components, to create a reverse vortex flow, counter to the direction of flow of gas through the stack and also along the periphery of the stack interior surface. Air from the first set of holes prevents recirculation zones from forming and the attendant accumulation of particulate deposits on the wall of the stack and will also return to the plenum any particulate swept up in the gas entering the stack. The second set of holes injects air in the same direction as the gas in the stack to compensate for the pressure drop and to prevent the concentration of condensate in the stack. A set of sprayers, receiving water from a second conduit, is located downstream of the second set of holes and sprays water into the gas to further cool it.

Combustor nozzles in gas turbine engines

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

Johnson, Thomas Edward; Keener, Christopher Paul; Stewart, Jason Thurman

2017-09-12

A micro-mixer nozzle for use in a combustor of a combustion turbine engine, the micro-mixer nozzle including: a fuel plenum defined by a shroud wall connecting a periphery of a forward tube sheet to a periphery of an aft tubesheet; a plurality of mixing tubes extending across the fuel plenum for mixing a supply of compressed air and fuel, each of the mixing tubes forming a passageway between an inlet formed through the forward tubesheet and an outlet formed through the aft tubesheet; and a wall mixing tube formed in the shroud wall.

Experimental and Computational Investigations of Plenum-to-Plenum Heat Transfer and Gas Dynamics under Natural Circulation in a Prismatic Very High Temperature Reactor

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

AL-Dahhan, Muthanna; Rizwan-Uddin, Rizwan; Usman, S.

All the goals and the objectives set for the project were successfully executed and achieved and all the milestones have been successfully completed. The results that have been obtained for the first time advance the scientific and engineering knowledge and understanding of the plenum-to plenum natural convection of prismatic block nuclear reactors that is encountered during accident or abnormal operation. These have been accomplished by developing and implementing for the first time unique and flexible scaled-down separate and integrated effects experimental plenumto- plenum facility (P2PF) with dual channels at this time that has been equipped with sophisticated measurement techniques integratedmore » in a novel way on the heated and cooled channels. The unique facility is an asset now that can be extended to research multiple channels and to study the effects of hot plumes in the plena for future projects if funding will be available. It can also be modified to research natural convection of pebble bed reactors. Hence, it complement the HTTF at Oregon State University. However, in this study, heat transfer coefficients from the inner wall surface to the flowing gas (both helium and air were used) and the radial temperature and gas velocity profiles have been measured and investigated along the height of the heated and cooled channels using in house developed wall flush mounted heat transfer probes, thermocouple with in house developed adjuster for radial movement with 1 mm increment inside the channel and hot wire anemometry with also in house developed adjuster for 1 mm radial movement inside the channel, respectively. Also advanced tracer technique has been developed to quantify also for the first time the dispersion of the gas dynamics of the hot and cold channels. The research has provided new knowledge and new benchmarking data that can be used to validate computational fluid dynamics (CFD) codes with conjugate heat transfer. The work and its results that have been performed within the budget have demonstrated their superior technical effectiveness and high economic feasibility to perform needed studies for safety analysis and assessment at least cost for these types of gas cooled very high temperature 4th generation nuclear reactors. Accordingly, the results obtained in this project and the unique facility and techniques that have been developed will benefit greatly the public by advancing the technology of the prismatic block nuclear reactors toward commercialization and to ensure they will be designed and operated safely by utilizing the obtained knowledge and having well validated CFD simulations integrated with heat transfer computations« less

Fission gas detection system

DOEpatents

Colburn, Richard P.

1985-01-01

A device for collecting fission gas released by a failed fuel rod which device uses a filter to pass coolant but which filter blocks fission gas bubbles which cannot pass through the filter due to the surface tension of the bubble.

Protection of extreme ultraviolet lithography masks. II. Showerhead flow mitigation of nanoscale particulate contamination [Protection of EUV lithography masks II: Showerhead flow mitigation of nanoscale particulate contamination

DOE PAGES

Klebanoff, Leonard E.; Torczynski, John R.; Geller, Anthony S.; ...

2015-03-27

An analysis is presented of a method to protect the reticle (mask) in an extreme ultraviolet (EUV) mask inspection tool using a showerhead plenum to provide a continuous flow of clean gas over the surface of a reticle. The reticle is suspended in an inverted fashion (face down) within a stage/holder that moves back and forth over the showerhead plenum as the reticle is inspected. It is essential that no particles of 10-nm diameter or larger be deposited on the reticle during inspection. Particles can originate from multiple sources in the system, and mask protection from each source is explicitlymore » analyzed. The showerhead plate has an internal plenum with a solid conical wall isolating the aperture. The upper and lower surfaces of the plate are thin flat sheets of porous-metal material. These porous sheets form the top and bottom showerheads that supply the region between the showerhead plate and the reticle and the region between the conical aperture and the Optics Zone box with continuous flows of clean gas. The model studies show that the top showerhead provides robust reticle protection from particles of 10-nm diameter or larger originating from the Reticle Zone and from plenum surfaces contaminated by exposure to the Reticle Zone. Protection is achieved with negligible effect on EUV transmission. Furthermore, the bottom showerhead efficiently protects the reticle from nanoscale particles originating from the Optics Zone.« less

Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

DOE PAGES

Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; ...

2014-10-12

The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertaintymore » in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.« less

A fission gas release correlation for uranium nitride fuel pins

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Davison, H. W.

1973-01-01

A model was developed to predict fission gas releases from UN fuel pins clad with various materials. The model was correlated with total release data obtained by different experimentors, over a range of fuel temperatures primarily between 1250 and 1660 K, and fuel burnups up to 4.6 percent. In the model, fission gas is transported by diffusion mechanisms to the grain boundaries where the volume grows and eventually interconnects with the outside surface of the fuel. The within grain diffusion coefficients are found from fission gas release rate data obtained using a sweep gas facility.

Fission gas release during power bumping at high burnup

NASA Astrophysics Data System (ADS)

Cunningham, M. E.; Freshley, M. D.; Lanning, D. D.

1993-03-01

Research to define the behavior of Zircaloy-clad light-water reactor fuel irradiated to high burnup levels was conducted by the High Burnup Effects Program (HBEP). One activity conducted by the HBEP was to "bump" the power level of irradiated, commercial light-water reactor fuel rods to design limit linear heat generation rates at end-of-life. These bumping irradiations simulated end-of-life design limit linear heat generation rates and provided data on the effects of short-term, high power irradiations at high burnup applicable to the design and operating constraints imposed by maximum allowable fuel rod internal gas pressure limits. Based on net fission gas release during the bumping irradiations, it was observed that higher burnup rods had greater rod-average fractional fission gas release than lower burnup rods at equal bumping powers. It was also observed that a hold period of 48 hours at the peak power was insufficient to achieve equilibrium fission gas release. Finally, differences in the prebump location of fission gas, i.e., within the UO 2 matrix or at grain boundaries, affected the fission gas release during the bumping irradiations.

PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

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

Pizzocri, D.; Rabiti, C.; Luzzi, L.

2016-09-01

This paper describes the development of a new numerical algorithm (called PolyPole-1) to efficiently solve the equation for intra-granular fission gas release in nuclear fuel. The work was carried out in collaboration with Politecnico di Milano and Institute for Transuranium Elements. The PolyPole-1 algorithms is being implemented in INL's fuels code BISON code as part of BISON's fission gas release model. The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of themore » corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this work, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, with the addition of polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of the PolyPole-1 solution is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work.« less

Effect of grain morphology on gas bubble swelling in UMo fuels – A 3D microstructure dependent Booth model

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

Hu, Shenyang; Burkes, Douglas; Lavender, Curt A.

2016-11-01

A three dimensional microstructure dependent swelling model is developed for studying the fission gas swelling kinetics in irradiated nuclear fuels. The model is extended from the Booth model [1] in order to investigate the effect of heterogeneous microstructures on gas bubble swelling kinetics. As an application of the model, the effect of grain morphology, fission gas diffusivity, and spatial dependent fission rate on swelling kinetics are simulated in UMo fuels. It is found that the decrease of grain size, the increase of grain aspect ratio for the grain having the same volume, and the increase of fission gas diffusivity (fissionmore » rate) cause the increase of swelling kinetics. Other heterogeneities such as second phases and spatial dependent thermodynamic properties including diffusivity of fission gas, sink and source strength of defects could be naturally integrated into the model to enhance the model capability.« less

Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

DOEpatents

DeVolpi, A.

1984-07-20

The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

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)

Recoil-α-fission and recoil-α-α-fission events observed in the reaction 48Ca + 243Am

NASA Astrophysics Data System (ADS)

Forsberg, U.; Rudolph, D.; Andersson, L.-L.; Di Nitto, A.; Düllmann, Ch. E.; Fahlander, C.; Gates, J. M.; Golubev, P.; Gregorich, K. E.; Gross, C. J.; Herzberg, R.-D.; Heßberger, F. P.; Khuyagbaatar, J.; Kratz, J. V.; Rykaczewski, K.; Sarmiento, L. G.; Schädel, M.; Yakushev, A.; Åberg, S.; Ackermann, D.; Block, M.; Brand, H.; Carlsson, B. G.; Cox, D.; Derkx, X.; Dobaczewski, J.; Eberhardt, K.; Even, J.; Gerl, J.; Jäger, E.; Kindler, B.; Krier, J.; Kojouharov, I.; Kurz, N.; Lommel, B.; Mistry, A.; Mokry, C.; Nazarewicz, W.; Nitsche, H.; Omtvedt, J. P.; Papadakis, P.; Ragnarsson, I.; Runke, J.; Schaffner, H.; Schausten, B.; Shi, Yue; Thörle-Pospiech, P.; Torres, T.; Traut, T.; Trautmann, N.; Türler, A.; Ward, A.; Ward, D. E.; Wiehl, N.

2016-09-01

Products of the fusion-evaporation reaction 48Ca + 243Am were studied with the TASISpec set-up at the gas-filled separator TASCA at the GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt, Germany. Amongst the detected thirty correlated α-decay chains associated with the production of element Z = 115, two recoil-α-fission and five recoil- α- α-fission events were observed. The latter five chains are similar to four such events reported from experiments performed at the Dubna gas-filled separator, and three such events reported from an experiment at the Berkeley gas-filled separator. The four chains observed at the Dubna gas-filled separator were assigned to start from the 2n-evaporation channel 289115 due to the fact that these recoil- α- α-fission events were observed only at low excitation energies. Contrary to this interpretation, we suggest that some of these recoil- α- α-fission decay chains, as well as some of the recoil- α- α-fission and recoil-α-fission decay chains reported from Berkeley and in this article, start from the 3n-evaporation channel 288115.

Neural net controlled tag gas sampling system for nuclear reactors

DOEpatents

Gross, Kenneth C.; Laug, Matthew T.; Lambert, John D. B.; Herzog, James P.

1997-01-01

A method and system for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod.

Modeling of Fission Gas Release in UO2

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

MH Krohn

2006-01-23

A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcadmore » [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].« less

On the effect of irradiation-induced resolution in modelling fission gas release in UO2 LWR fuel

NASA Astrophysics Data System (ADS)

Lösönen, Pekka

2017-12-01

Irradiation resolution of gas atoms and vacancies from intra- and intergranular bubbles in sintered UO2 fuel was studied by comparing macroscopic models with a more mechanistic approach. The applied macroscopic models imply the resolution rate of gas atoms to be proportional to gas concentration in intragranular bubbles and at grain boundary (including intergranular bubbles). A relation was established between the macroscopic models and a single encounter of an energetic fission fragment with a bubble. The effect of bubble size on resolution was quantified. The number of resoluted gas atoms per encounter of a fission fragment per bubble was of the same order of magnitude for intra- and intergranular bubbles. However, the resulting macroscopic resolution rate of gas atoms was about two orders of magnitude larger from intragranular bubbles. The number of vacancies resoluted from a grain face bubble by a passing fission fragment was calculated. The obtained correlations for resolution of gas atoms from intragranular bubbles and grain boundaries and for resolution of vacancies from grain face bubbles were used to demonstrate the effect of irradiation resolution on fission gas release.

First-principles study of fission gas incorporation and migration in zirconium nitride

DOE PAGES

Mei, Zhi-Gang; Liang, Linyun; Yacout, Abdellatif M.

2017-03-24

To evaluate the effectiveness of ZrN as a diffusion barrier against fission gases, we investigate in this paper the incorporation and migration of fission gas atoms, with a focus on Xe, in ZrN by first-principles calculations. The formations of point defects in ZrN, including vacancies, interstitials, divacancies, Frenkel pairs, and Schottky defects, are first studied. Among all the defects, the Schottky defect with two vacancies as first nearest neighbor is predicted to be the most favorable incorporation site for fission gas Xe in ZrN. The migration of Xe gas atom in ZrN is investigated through two diffusion mechanisms, i.e., interstitialmore » and vacancy-assisted diffusions. The migration barrier of Xe gas atom through the intrinsic interstitials in ZrN is considerably lower than that through vacancies. Finally, therefore, at low temperatures fission gas Xe atoms diffuse mainly through interstitials in single crystal ZrN, whereas at high temperatures Xe may diffuse in ZrN assisted by vacancies.« less

Modified laminar flow biological safety cabinet.

PubMed

McGarrity, G J; Coriell, L L

1974-10-01

Tests are reported on a modified laminar flow biological safety cabinet in which the return air plenum that conducts air from the work area to the high efficiency particulate air filters is under negative pressure. Freon gas released inside the cabinet could not be detected outside by a freon gas detection method capable of detecting 10(-6) cc/s. When T3 bacteriophage was aerosolized 5 cm outside the front opening in 11 tests, no phage could be detected inside the cabinet with the motor-filter unit in operation. An average of 2.8 x 10(5) plaque-forming units (PFU)/ft(3) (ca. 0.028 m(3)) were detected with the motor-filter unit not in operation, a penetration of 0.0%. Aerosolization 5 cm inside the cabinet yielded an average of 10 PFU/ft(3) outside the cabinet with the motor-filter unit in operation and an average of 4.1 x 10(5) PFU/ft(3) with the motor-filter unit not in operation, a penetration of 0.002%. These values are the same order of effectiveness as the positive-pressure laminar flow biological safety cabinets previously tested. The advantages of the negative-pressure return plenum design include: (i) assurance that if cracks or leaks develop in the plenum it will not lead to discharge of contaminated air into the laboratory; and (ii) the price is lower due to reduced manufacturing costs.

Dish/stirling hybrid-receiver

DOEpatents

Mehos, Mark S.; Anselmo, Kenneth M.; Moreno, James B.; Andraka, Charles E.; Rawlinson, K. Scott; Corey, John; Bohn, Mark S.

2002-01-01

A hybrid high-temperature solar receiver is provided which comprises a solar heat-pipe-receiver including a front dome having a solar absorber surface for receiving concentrated solar energy, a heat pipe wick, a rear dome, a sidewall joining the front and the rear dome, and a vapor and a return liquid tube connecting to an engine, and a fossil fuel fired combustion system in radial integration with the sidewall for simultaneous operation with the solar heat pipe receiver, the combustion system comprising an air and fuel pre-mixer, an outer cooling jacket for tangentially introducing and cooling the mixture, a recuperator for preheating the mixture, a burner plenum having an inner and an outer wall, a porous cylindrical metal matrix burner firing radially inward facing a sodium vapor sink, the mixture ignited downstream of the matrix forming combustion products, an exhaust plenum, a fossil-fuel heat-input surface having an outer surface covered with a pin-fin array, the combustion products flowing through the array to give up additional heat to the receiver, and an inner surface covered with an extension of the heat-pipe wick, a pin-fin shroud sealed to the burner and exhaust plenums, an end seal, a flue-gas diversion tube and a flue-gas valve for use at off-design conditions to limit the temperature of the pre-heated air and fuel mixture, preventing pre-ignition.

FASTGRASS implementation in BISON and Fission gas behavior characterization in UO 2 and connection to validating MARMOT

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

Yun, Di; Mo, Kun; Ye, Bei

2015-09-30

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO 2. Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module inmore » the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO 2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.« less

Unit mechanisms of fission gas release: Current understanding and future needs

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

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel properties and, once the gas is released into the gap between the fuel and cladding, lowering gap thermal conductivity and increasing gap pressure. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are being applied to provide unprecedented understanding of the unit mechanisms that define the fission product behavior. In this article, existing research on the basic mechanisms behind the various stages of fission gas releasemore » during normal reactor operation are summarized and critical areas where experimental and simulation work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior during reactor operation and to design fuels that have improved fission product retention. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

DOE PAGES

Tonks, Michael; Andersson, David; Devanathan, Ram; ...

2018-03-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

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

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

NASA Astrophysics Data System (ADS)

Tonks, Michael; Andersson, David; Devanathan, Ram; Dubourg, Roland; El-Azab, Anter; Freyss, Michel; Iglesias, Fernando; Kulacsy, Katalin; Pastore, Giovanni; Phillpot, Simon R.; Welland, Michael

2018-06-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

Serial cooling of a combustor for a gas turbine engine

DOEpatents

Abreu, Mario E.; Kielczyk, Janusz J.

2001-01-01

A combustor for a gas turbine engine uses compressed air to cool a combustor liner and uses at least a portion of the same compressed air for combustion air. A flow diverting mechanism regulates compressed air flow entering a combustion air plenum feeding combustion air to a plurality of fuel nozzles. The flow diverting mechanism adjusts combustion air according to engine loading.

Neural net controlled tag gas sampling system for nuclear reactors

DOEpatents

Gross, K.C.; Laug, M.T.; Lambert, J.B.; Herzog, J.P.

1997-02-11

A method and system are disclosed for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod. 12 figs.

Pulse detonation engines and components thereof

NASA Technical Reports Server (NTRS)

Tangirala, Venkat Eswarlu (Inventor); Rasheed, Adam (Inventor); Vandervort, Christian Lee (Inventor); Dean, Anthony John (Inventor)

2009-01-01

A pulse detonation engine comprises a primary air inlet; a primary air plenum located in fluid communication with the primary air inlet; a secondary air inlet; a secondary air plenum located in fluid communication with the secondary air inlet, wherein the secondary air plenum is substantially isolated from the primary air plenum; a pulse detonation combustor comprising a pulse detonation chamber, wherein the pulse detonation chamber is located downstream of and in fluid communication with the primary air plenum; a coaxial liner surrounding the pulse detonation combustor defining a cooling plenum, wherein the cooling plenum is in fluid communication with the secondary air plenum; an axial turbine assembly located downstream of and in fluid communication with the pulse detonation combustor and the cooling plenum; and a housing encasing the primary air plenum, the secondary air plenum, the pulse detonation combustor, the coaxial liner, and the axial turbine assembly.

Regeneratively cooled coal combustor/gasifier with integral dry ash removal

DOEpatents

Beaufrere, Albert H.

1983-10-04

A coal combustor/gasifier is disclosed which produces a low or medium combustion gas for further combustion in modified oil or gas fired furnaces or boilers. Two concentric shells define a combustion volume within the inner shell and a plenum between them through which combustion air flows to provide regenerative cooling of the inner shell for dry ash operation. A fuel flow and a combustion air flow having opposed swirls are mixed and burned in a mixing-combustion portion of the combustion volume and the ash laden combustion products flow with a residual swirl into an ash separation region. The ash is cooled below the fusion temperature and is moved to the wall by centrifugal force where it is entrained in the cool wall boundary layer. The boundary layer is stabilized against ash re-entrainment as it is moved to an ash removal annulus by a flow of air from the plenum through slots in the inner shell, and by suction on an ash removal skimmer slot.

A fast rise-rate, adjustable-mass-bit gas puff valve for energetic pulsed plasma experiments

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

Loebner, Keith T. K., E-mail: kloebner@stanford.edu; Underwood, Thomas C.; Cappelli, Mark A.

2015-06-15

A fast rise-rate, variable mass-bit gas puff valve based on the diamagnetic repulsion principle was designed, built, and experimentally characterized. The ability to hold the pressure rise-rate nearly constant while varying the total overall mass bit was achieved via a movable mechanical restrictor that is accessible while the valve is assembled and pressurized. The rise-rates and mass-bits were measured via piezoelectric pressure transducers for plenum pressures between 10 and 40 psig and restrictor positions of 0.02-1.33 cm from the bottom of the linear restrictor travel. The mass-bits were found to vary linearly with the restrictor position at a given plenummore » pressure, while rise-rates varied linearly with plenum pressure but exhibited low variation over the range of possible restrictor positions. The ability to change the operating regime of a pulsed coaxial plasma deflagration accelerator by means of altering the valve parameters is demonstrated.« less

Fission-gas release from uranium nitride at high fission rate density

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Kirchgessner, T. A.; Tambling, T. N.

1973-01-01

A sweep gas facility has been used to measure the release rates of radioactive fission gases from small UN specimens irradiated to 8-percent burnup at high fission-rate densities. The measured release rates have been correlated with an equation whose terms correspond to direct recoil release, fission-enhanced diffusion, and atomic diffusion (a function of temperature). Release rates were found to increase linearly with burnups between 1.5 and 8 percent. Pore migration was observed after operation at 1550 K to over 6 percent burnup.

Oil cooled, hermetic refrigerant compressor

DOEpatents

English, William A.; Young, Robert R.

1985-01-01

A hermetic refrigerant compressor having an electric motor and compressor assembly in a hermetic shell is cooled by oil which is first cooled in an external cooler 18 and is then delivered through the shell to the top of the motor rotor 24 where most of it is flung radially outwardly within the confined space provided by the cap 50 which channels the flow of most of the oil around the top of the stator 26 and then out to a multiplicity of holes 52 to flow down to the sump and provide further cooling of the motor and compressor. Part of the oil descends internally of the motor to the annular chamber 58 to provide oil cooling of the lower part of the motor, with this oil exiting through vent hole 62 also to the sump. Suction gas with entrained oil and liquid refrigerant therein is delivered to an oil separator 68 from which the suction gas passes by a confined path in pipe 66 to the suction plenum 64 and the separated oil drops from the separator to the sump. By providing the oil cooling of the parts, the suction gas is not used for cooling purposes and accordingly increase in superheat is substantially avoided in the passage of the suction gas through the shell to the suction plenum 64.

Oil cooled, hermetic refrigerant compressor

DOEpatents

English, W.A.; Young, R.R.

1985-05-14

A hermetic refrigerant compressor having an electric motor and compressor assembly in a hermetic shell is cooled by oil which is first cooled in an external cooler and is then delivered through the shell to the top of the motor rotor where most of it is flung radially outwardly within the confined space provided by the cap which channels the flow of most of the oil around the top of the stator and then out to a multiplicity of holes to flow down to the sump and provide further cooling of the motor and compressor. Part of the oil descends internally of the motor to the annular chamber to provide oil cooling of the lower part of the motor, with this oil exiting through vent hole also to the sump. Suction gas with entrained oil and liquid refrigerant therein is delivered to an oil separator from which the suction gas passes by a confined path in pipe to the suction plenum and the separated oil drops from the separator to the sump. By providing the oil cooling of the parts, the suction gas is not used for cooling purposes and accordingly increase in superheat is substantially avoided in the passage of the suction gas through the shell to the suction plenum. 3 figs.

Fission-gas-release rates from irradiated uranium nitride specimens

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Kirchgessner, T. A.; Tambling, T. N.

1973-01-01

Fission-gas-release rates from two 93 percent dense UN specimens were measured using a sweep gas facility. Specimen burnup rates averaged .0045 and .0032 percent/hr, and the specimen temperatures ranged from 425 to 1323 K and from 552 to 1502 K, respectively. Burnups up to 7.8 percent were achieved. Fission-gas-release rates first decreased then increased with burnup. Extensive interconnected intergranular porosity formed in the specimen operated at over 1500 K. Release rate variation with both burnup and temperature agreed with previous irradiation test results.

Development of the Los Alamos continuous high average-power microsecond pulser ion accelerator

NASA Astrophysics Data System (ADS)

Bitteker, L. J.; Wood, B. P.; Davis, H. A.; Waganaar, W. J.; Boyd, I. D.; Lovberg, R. H.

2000-10-01

The continuous high average-power microsecond pulser (CHAMP) ion accelerator is being constructed at Los Alamos National Laboratory. Progress on the testing of the CHAMP diode is discussed. A direct simulation Monte Carlo computer code is used to investigate the puffed gas fill of the CHAMP anode. High plenum pressures and low plenum volumes are found to be desirable for effective gas puffs. The typical gas fill time is 150-180 μs from initiation of valve operation to end of fill. Results of anode plasma production at three stages of development are discussed. Plasma properties are monitored with electric and magnetic field probes. From this data, the near coil plasma density under nominal conditions is found to be on the order of 1×1016 cm-3. Large error is associated with this calculation due to inconsistencies between tests and the limitations of the instrumentation used. The diode insulating magnetic field is observed to result in lower density plasma with a more diffuse structure than for the cases when the insulating field is not applied. The importance of these differences in plasma quality on the beam production is yet to be determined.

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)

High Burnup Effects Program

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

Barner, J.O.; Cunningham, M.E.; Freshley, M.D.

1990-04-01

This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the coursemore » of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs.« less

Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

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

Miao, Yinbin; Ye, Bei; Hofman, Gerard

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U 3Si 2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U 3Si 2 at LWR conditions. The fission gas behavior of U 3Si 2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranularmore » bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U 3Si 2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U 3Si 2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U 3Si 2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.« less

Derivation of effective fission gas diffusivities in UO2 from lower length scale simulations and implementation of fission gas diffusion models in BISON

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

Andersson, Anders David Ragnar; Pastore, Giovanni; Liu, Xiang-Yang

2014-11-07

This report summarizes the development of new fission gas diffusion models from lower length scale simulations and assessment of these models in terms of annealing experiments and fission gas release simulations using the BISON fuel performance code. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations, continuum models for diffusion of xenon (Xe) in UO 2 were derived for both intrinsic conditions and under irradiation. The importance of the large X eU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequencemore » of its high mobility and stability. These models were implemented in the MARMOT phase field code, which is used to calculate effective Xe diffusivities for various irradiation conditions. The effective diffusivities were used in BISON to calculate fission gas release for a number of test cases. The results are assessed against experimental data and future directions for research are outlined based on the conclusions.« less

A model describing intra-granular fission gas behaviour in oxide fuel for advanced engineering tools

NASA Astrophysics Data System (ADS)

Pizzocri, D.; Pastore, G.; Barani, T.; Magni, A.; Luzzi, L.; Van Uffelen, P.; Pitts, S. A.; Alfonsi, A.; Hales, J. D.

2018-04-01

The description of intra-granular fission gas behaviour is a fundamental part of any model for the prediction of fission gas release and swelling in nuclear fuel. In this work we present a model describing the evolution of intra-granular fission gas bubbles in terms of bubble number density and average size, coupled to gas release to grain boundaries. The model considers the fundamental processes of single gas atom diffusion, gas bubble nucleation, re-solution and gas atom trapping at bubbles. The model is derived from a detailed cluster dynamics formulation, yet it consists of only three differential equations in its final form; hence, it can be efficiently applied in engineering fuel performance codes while retaining a physical basis. We discuss improvements relative to previous single-size models for intra-granular bubble evolution. We validate the model against experimental data, both in terms of bubble number density and average bubble radius. Lastly, we perform an uncertainty and sensitivity analysis by propagating the uncertainties in the parameters to model results.

Monitoring the Gas Composition of the NIFFTE Time Projection Chamber

NASA Astrophysics Data System (ADS)

Towell, Travis; Travis Towell Collaboration

2017-09-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) at Los Alamos National Laboratory(LANL) is using a Time Projection Chamber (TPC) to measure with high precision the cross section ratio of U238 to P239. When the neutron beam hits a target, it may emit fission fragments. As the fission fragments travels through the chamber, it ionizes the gas it passes through. Based on the time it takes for the ions to drift to the pad planes and the hit location of the ions, the path of fission fragments can be determined. Knowing the composition of the gas mixture is vital to accurately reconstruct the data. A Binary Gas Analyzer (BGA) is used to measure the gas composition. To confirm the accuracy of the BGA, varying amounts of nitrogen and carbon dioxide were flowed through a test gas system. Several tests were performed to validate that the BGA for our gas system is working properly. This poster will describe the test gas system setup, tests of the BGA, and elaborate on the main goals of the NIFFTE experiment.

Cooling using complimentary tapered plenums

DOEpatents

Hall, Shawn Anthony [Pleasantville, NY

2006-08-01

Where a fluid cooling medium cools a plurality of heat-producing devices arranged in a row along a generalized coordinate direction, with a space between each adjacent pair of devices, each space may have a partition that defines a boundary between a first plenum and a second plenum. The first plenum carries cooling medium across an entrance and thence into a first heat-producing device located on a first side of the partition facing the first plenum. The second plenum carries cooling medium away from a second heat-producing device located on a second side of the partition facing the second plenum and thence across an exit. The partition is disposed so that the first plenum becomes smaller in cross-sectional area as distance increases from the entrance, and the second plenum becomes larger in cross sectional area as distance decreases toward the exit.

Characterization of fission gas bubbles in irradiated U-10Mo fuel

DOE PAGES

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; ...

2017-06-06

A simple, repeatable method for characterization of fission gas bubbles in irradiated U-Mo fuels has been developed. This method involves mechanical potting and polishing of samples along with examination with a scanning electron microscope located outside of a hot cell. The commercially available software packages CellProfiler, MATLAB, and Mathematica are used to segment and analyze the captured images. The results are compared and contrasted. Finally, baseline methods for fission gas bubble characterization are suggested for consideration and further development.

Gas turbine engines with particle traps

DOEpatents

Boyd, Gary L.; Sumner, D. Warren; Sheoran, Yogendra; Judd, Z. Daniel

1992-01-01

A gas turbine engine (10) incorporates a particle trap (46) that forms an entrapment region (73) in a plenum (24) which extends from within the combustor (18) to the inlet (32) of a radial-inflow turbine (52, 54). The engine (10) is thereby adapted to entrap particles that originate downstream from the compressor (14) and are otherwise propelled by combustion gas (22) into the turbine (52, 54). Carbonaceous particles that are dislodged from the inner wall (50) of the combustor (18) are incinerated within the entrapment region (73) during operation of the engine (10).

Ventilation for an enclosure of a gas turbine and related method

DOEpatents

Schroeder, Troy Joseph; Leach, David; O'Toole, Michael Anthony

2002-01-01

A ventilation scheme for a rotary machine supported on pedestals within an enclosure having a roof, end walls and side walls with the machine arranged parallel to the side walls, includes ventilation air inlets located in a first end wall of the enclosure; a barrier wall located within the enclosure, proximate the first end wall to thereby create a plenum chamber. The barrier wall is constructed to provide a substantially annular gap between the barrier wall and a casing of the turbine to thereby direct ventilation air axially along the turbine; one or more ventilation air outlets located proximate a second, opposite end wall on the roof of the enclosure. In addition, one or more fans are provided for pulling ventilating air into said plenum chamber via the ventilation air inlets.

Plenum response to simulated disturbances of the model and fan inlet guide vanes in a transonic tunnel

NASA Technical Reports Server (NTRS)

Gloss, B. B.

1980-01-01

In order to aid in the design of the National Transonic Facility (NTF) control system, test section/plenum response studies were carried out in a 0.186 scale model of the NTF high speed duct. Two types of disturbances, those induced by the model and those induced by the compressor inlet guide vanes were simulated. Some observations with regard to the test section/plenum response tests are summarized as follows. A resonance frequency for the test section/plenum area of the tunnel of approximately 50 Hz was observed for Mach numbers from 0.40 to 0.90. However, since the plenum is 3.1 times (based on volume) too large for the scaled size of the test section, care must be taken in extrapolating these data to NTF conditions. The plenum pressure data indicate the existence of pressure gradients in the plenum. The test results indicate that the difference between test section static pressure and plenum pressure is dependent on test section flow conditions. Plenum response to inlet guide vane type disturbances appears to be slower than plenum response to test section disturbances.

Lateral Reaction Jet Flow Interaction Effects on a Generic Fin-Stabilized Munition in Supersonic Crossflows

DTIC Science & Technology

2013-11-01

freestream conditions ( 0 =300 K). .........22  Table 7. Results from nozzle parameter study, variation with jet gas total temperature (AR=1, M=2.5...end. Two additional supersonic nozzles of AR=2 and AR=8 (figures 3e and 3f) were also investigated, also with a throat diameter of 2.54 mm. The...walls, due to the different flow properties from the gas expansion there. Therefore, the plenum and nozzle exit walls were modeled with an advanced

Coupling Nuclear Induced Phonon Propagation with Conversion Electron Moessbauer Spectroscopy

DTIC Science & Technology

2015-06-18

penetrating and a small detector with low density will be insensitive to the gammas. Thus, a small gas proportional counter is ideal for this application ...and Materials Science Vol. 1. Plenum: New York, 1993. 16. Long, G. J. and Stevens, J. G., eds. (1986) " Industrial applications of the Mössbauer...proportional gas detector attached to left side of 1” diameter stainless steel type-310 bar, phonon source encased in a mounting device attached to the right

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)

Irradiation behavior of U 6Mn-Al dispersion fuel elements

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

2000-02-01

Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

Effects of injection nozzle exit width on rotating detonation engine

NASA Astrophysics Data System (ADS)

Sun, Jian; Zhou, Jin; Liu, Shijie; Lin, Zhiyong; Cai, Jianhua

2017-11-01

A series of numerical simulations of RDE modeling real injection nozzles with different exit widths are performed in this paper. The effects of nozzle exit width on chamber inlet state, plenum flowfield and detonation propagation are analyzed. The results are compared with that using an ideal injection model. Although the ideal injection model is a good approximation method to model RDE inlet, the two-dimensional effects of real nozzles are ignored in the ideal injection model so that some complicated phenomena such as the reflected waves caused by the nozzle walls and the reversed flow into the nozzles can not be modeled accurately. Additionally, the ideal injection model overpredicts the block ratio. In all the cases that stabilize at one-wave mode, the block ratio increases as the nozzle exit width gets smaller. The dual-wave mode case also has a relatively high block ratio. A pressure oscillation in the plenum with the same main frequency with the rotating detonation wave is observed. A parameter σ is applied to describe the non-uniformity in the plenum. σ increases as the nozzle exit width gets larger. Under some condition, the heat release on the interface of fresh premixed gas layer and detonation products can be strong enough to induce a new detonation wave. A spontaneous mode-transition process is observed for the smallest exit width case. Due to the detonation products existing in the premixed gas layer before the detonation wave, the detonation wave will propagate through reactants and products alternately, and therefore its strength will vary with time, especially near the chamber inlet. This tendency gets weaker as the injection nozzle exit width increases.

Flow design and simulation of a gas compression system for hydrogen fusion energy production

NASA Astrophysics Data System (ADS)

Avital, E. J.; Salvatore, E.; Munjiza, A.; Suponitsky, V.; Plant, D.; Laberge, M.

2017-08-01

An innovative gas compression system is proposed and computationally researched to achieve a short time response as needed in engineering applications such as hydrogen fusion energy reactors and high speed hammers. The system consists of a reservoir containing high pressure gas connected to a straight tube which in turn is connected to a spherical duct, where at the sphere’s centre plasma resides in the case of a fusion reactor. Diaphragm located inside the straight tube separates the reservoir’s high pressure gas from the rest of the plenum. Once the diaphragm is breached the high pressure gas enters the plenum to drive pistons located on the inner wall of the spherical duct that will eventually end compressing the plasma. Quasi-1D and axisymmetric flow formulations are used to design and analyse the flow dynamics. A spike is designed for the interface between the straight tube and the spherical duct to provide a smooth geometry transition for the flow. Flow simulations show high supersonic flow hitting the end of the spherical duct, generating a return shock wave propagating upstream and raising the pressure above the reservoir pressure as in the hammer wave problem, potentially giving temporary pressure boost to the pistons. Good agreement is revealed between the two flow formulations pointing to the usefulness of the quasi-1D formulation as a rapid solver. Nevertheless, a mild time delay in the axisymmetric flow simulation occurred due to moderate two-dimensionality effects. The compression system is settled down in a few milliseconds for a spherical duct of 0.8 m diameter using Helium gas and a uniform duct cross-section area. Various system geometries are analysed using instantaneous and time history flow plots.

Effects of Geometric Variations on Lift Augmentation of Simple-plenum-chamber Ground-effect Models

NASA Technical Reports Server (NTRS)

Davenport, Edwin E.

1961-01-01

Considerable interest has been shown during recent years in ground-effect vehicles. Of the various types proposed, the simple-plenum-chamber vehicle has indicated promise because, although the lift augmentation obtainable appears to be less than that of an annular jet, it may be somewhat less complicated structurally. The present investigation was undertaken to study the effects of some geometric variations upon lift augmentation of a simple plenum chamber within ground proximity. The variables included the ratio inlet area to exit area, plenum-chamber depth, and entrance configuration. An optimum plenum-chamber depth appeared to be between 3 and 10 percent of the plenum-chamber diameter with a ratio of inlet diameter to plenum-chamber diameter of 0.15 for the range of plenum-chamber depths investigated. The most important effect of multiple inlets was the elimination of negative lift augmentation, which was experienced with single sharp-edged inlets, at intermediate heights. Installation of a flared inlet and a turning-vane assembly improved lift augmentation of a single-inlet configuration at intermediate heights.

Combustion system for hybrid solar fossil fuel receiver

DOEpatents

Mehos, Mark S.; Anselmo, Kenneth M.; Moreno, James B.; Andraka, Charles E.; Rawlinson, K. Scott; Corey, John; Bohn, Mark S.

2004-05-25

A combustion system for a hybrid solar receiver comprises a pre-mixer which combines air and fuel to form an air-fuel mixture. The mixture is introduced tangentially into a cooling jacket. A burner plenum is fluidically connected to the cooling jacket such that the burner plenum and the cooling jacket are arranged in thermal contact with one another. The air-fuel mixture flows through the cooling jacket cooling the burner plenum to reduce pre-ignition of the air-fuel mixture in the burner plenum. A combustion chamber is operatively associated with and open to the burner plenum to receive the air-fuel mixture from the burner plenum. An igniter is operatively positioned in the combustion chamber to combust the air-fuel mixture, releasing heat. A recuperator is operatively associated with the burner plenum and the combustion chamber and pre-heats the air-fuel mixture in the burner plenum with heat from the combustion chamber. A heat-exchanger is operatively associated and in thermal contact with the combustion chamber. The heat-exchanger provides heat for the hybrid solar receiver.

Effects of Noise of Offshore Oil and Gas Operations on Marine Mammals - An Introductory Assessment. Volume 2

DTIC Science & Technology

1982-09-01

p 110-141, 1971. 61 Penner, RH and J Kadane, Tursiops Biosonar Detection in Noise. In: Animal Sonar Systems. RF Busnel and JF Fish, eds, p 957-959...and J Kadane, Tursiops Biosonar Detection in Noise, In: Animal Sonar .* Systems, RF Busnel and JF Fish, eds, p 957-959, Plenum Press, 1980, 62

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.

Premixed direct injection disk

DOEpatents

York, William David; Ziminsky, Willy Steve; Johnson, Thomas Edward; Lacy, Benjamin; Zuo, Baifang; Uhm, Jong Ho

2013-04-23

A fuel/air mixing disk for use in a fuel/air mixing combustor assembly is provided. The disk includes a first face, a second face, and at least one fuel plenum disposed therebetween. A plurality of fuel/air mixing tubes extend through the pre-mixing disk, each mixing tube including an outer tube wall extending axially along a tube axis and in fluid communication with the at least one fuel plenum. At least a portion of the plurality of fuel/air mixing tubes further includes at least one fuel injection hole have a fuel injection hole diameter extending through said outer tube wall, the fuel injection hole having an injection angle relative to the tube axis. The invention provides good fuel air mixing with low combustion generated NOx and low flow pressure loss translating to a high gas turbine efficiency, that is durable, and resistant to flame holding and flash back.

Fission Xenon on Mars

NASA Technical Reports Server (NTRS)

Mathew, K. J.; Marti, K.; Marty, B.

2002-01-01

Fission Xe components due to Pu-244 decay in the early history of Mars have been identified in nakhlites; as in the case of ALH84001 and Chassigny the fission gas was assimilated into indigenous solar-type Xe. Additional information is contained in the original extended abstract.

Mechanistic prediction of fission-gas behavior during in-cell transient heating tests on LWR fuel using the GRASS-SST and FASTGRASS computer codes

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

Rest, J; Gehl, S M

1979-01-01

GRASS-SST and FASTGRASS are mechanistic computer codes for predicting fission-gas behavior in UO/sub 2/-base fuels during steady-state and transient conditions. FASTGRASS was developed in order to satisfy the need for a fast-running alternative to GRASS-SST. Althrough based on GRASS-SST, FASTGRASS is approximately an order of magnitude quicker in execution. The GRASS-SST transient analysis has evolved through comparisons of code predictions with the fission-gas release and physical phenomena that occur during reactor operation and transient direct-electrical-heating (DEH) testing of irradiated light-water reactor fuel. The FASTGRASS calculational procedure is described in this paper, along with models of key physical processes included inmore » both FASTGRASS and GRASS-SST. Predictions of fission-gas release obtained from GRASS-SST and FASTGRASS analyses are compared with experimental observations from a series of DEH tests. The major conclusions is that the computer codes should include an improved model for the evolution of the grain-edge porosity.« less

Systems and methods for preventing flashback in a combustor assembly

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

Johnson, Thomas Edward; Ziminsky, Willy Steve; Stevenson, Christian Xavier

2016-04-05

Embodiments of the present application include a combustor assembly. The combustor assembly may include a combustion chamber, a first plenum, a second plenum, and one or more elongate air/fuel premixing injection tubes. Each of the elongate air/fuel premixing injection tubes may include a first length at least partially disposed within the first plenum and configured to receive a first fluid from the first plenum. Moreover, each of the elongate air/fuel premixing injection tubes may include a second length disposed downstream of the first length and at least partially disposed within the second plenum. The second length may be formed ofmore » a porous wall configured to allow a second fluid from the second plenum to enter the second length and create a boundary layer about the porous wall.« less

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)

Possible Effects of Noise from Offshore Oil and Gas Drilling Activities on Marine Mammals: A Survey of the Literature

DTIC Science & Technology

1982-01-01

in Baleen Whales, New York Acad Sci, 188, p 110-141, 1971. 61 Penner, RI and J Kadane, Tursiops Biosonar Detection in Noise, In: Animal Sonar Systems... Biosonar Detection in Noise, In: Animal Sonar Systems, RF Busnel and JF Fish, eds, p 957-959, Plenum Press, 1980. 62. Nishiwake, M and A Sasao, Human

METHOD OF OPERATING A NEUTRONIC REACTOR

DOEpatents

Turkevich, A.

1963-01-22

This patent relates to one step in a method of operating a neutronic reactor consisting of a slurry of fissionable material in heavy water. Deuterium gas is passed through the slurry to sweep gaseous fission products therefrom and the deuterium is then separated from the gaseous fission products. (AEC)

The 3-D CFD modeling of gas turbine combustor-integral bleed flow interaction

NASA Technical Reports Server (NTRS)

Chen, D. Y.; Reynolds, R. S.

1993-01-01

An advanced 3-D Computational Fluid Dynamics (CFD) model was developed to analyze the flow interaction between a gas turbine combustor and an integral bleed plenum. In this model, the elliptic governing equations of continuity, momentum and the k-e turbulence model were solved on a boundary-fitted, curvilinear, orthogonal grid system. The model was first validated against test data from public literature and then applied to a gas turbine combustor with integral bleed. The model predictions agreed well with data from combustor rig testing. The model predictions also indicated strong flow interaction between the combustor and the integral bleed. Integral bleed flow distribution was found to have a great effect on the pressure distribution around the gas turbine combustor.

Counterflow heat exchanger with core and plenums at both ends

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

Bejan, A.; Alalaimi, M.; Lorente, S.

2016-04-22

Here, this paper illustrates the morphing of flow architecture toward greater performance in a counterflow heat exchanger. The architecture consists of two plenums with a core of counterflow channels between them. Each stream enters one plenum and then flows in a channel that travels the core and crosses the second plenum. The volume of the heat exchanger is fixed while the volume fraction occupied by each plenum is variable. Performance is driven by two objectives, simultaneously: low flow resistance and low thermal resistance. The analytical and numerical results show that the overall flow resistance is the lowest when the coremore » is absent, and each plenum occupies half of the available volume and is oriented in counterflow with the other plenum. In this configuration, the thermal resistance also reaches its lowest value. These conclusions hold for fully developed laminar flow and turbulent flow through the core. The curve for effectiveness vs number of heat transfer units (N tu) is steeper (when N tu < 1) than the classical curves for counterflow and crossflow.« less

Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system

NASA Astrophysics Data System (ADS)

Sobolev, V.; Lemehov, S.; Messaoudi, N.; Van Uffelen, P.; Aı̈t Abderrahim, H.

2003-06-01

The Belgian Nuclear Research Centre, SCK • CEN, is currently working on the pre-design of the multipurpose accelerator-driven system (ADS) MYRRHA. A demonstration of the possibility of transmutation of minor actinides and long-lived fission products with a realistic design of experimental fuel targets and prognosis of their behaviour under typical ADS conditions is an important task in the MYRRHA project. In the present article, the irradiation behaviour of three different oxide fuel mixtures, containing americium and plutonium - (Am,Pu,U)O 2- x with urania matrix, (Am,Pu,Th)O 2- x with thoria matrix and (Am,Y,Pu,Zr)O 2- x with inert zirconia matrix stabilised by yttria - were simulated with the new fuel performance code MACROS, which is under development and testing at the SCK • CEN. All the fuel rods were considered to be of the same design and sizes: annular fuel pellets, helium bounded with the stainless steel cladding, and a large gas plenum. The liquid lead-bismuth eutectic was used as coolant. Typical irradiation conditions of the hottest fuel assembly of the MYRRHA subcritical core were pre-calculated with the MCNPX code and used in the following calculations as the input data. The results of prediction of the thermo-mechanical behaviour of the designed rods with the considered fuels during three irradiation cycles of 90 EFPD are presented and discussed.

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

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

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Fuel injection assembly for use in turbine engines and method of assembling same

DOEpatents

Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

2015-12-15

A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

Nuclear Power from Fission Reactors. An Introduction.

ERIC Educational Resources Information Center

Department of Energy, Washington, DC. Technical Information Center.

The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

NASA Astrophysics Data System (ADS)

Johnson, Lawrence; Ferry, Cécile; Poinssot, Christophe; Lovera, Patrick

2005-11-01

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO 2 or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.

On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

NASA Astrophysics Data System (ADS)

Restani, R.; Horvath, M.; Goll, W.; Bertsch, J.; Gavillet, D.; Hermann, A.; Martin, M.; Walker, C. T.

2016-12-01

Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

NASA Astrophysics Data System (ADS)

Lemoine, F.

1997-09-01

Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning the transient fission gas behavior and the clad loading mechanisms. The importance of the rim zone is demonstrated based on three experiments resulting in clad failure at low enthalpy, which are explained by energetic considerations. High gas release in non-failure tests with low energy deposition underlines the importance of grain boundary and porosity gas. Measured final releases are strongly correlated to the microstructure evolution, depending on energy deposition, pulse width, initial and refabricated fuel rod design. Observed helium release can also increase internal pressure and gives hints to the gas behavior understanding.

Fission gas bubble identification using MATLAB's image processing toolbox

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

Collette, R.

Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. This study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding proved to bemore » the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods. - Highlights: •Automated image processing can aid in the fuel qualification process. •Routines are developed to characterize fission gas bubbles in irradiated U–Mo fuel. •Frequency domain filtration effectively eliminates FIB curtaining artifacts. •Adaptive thresholding proved to be the most accurate segmentation method. •The techniques established are ready to be applied to large scale data extraction testing.« less

Fission gas bubble identification using MATLAB's image processing toolbox

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

Collette, R.; King, J.; Keiser, Jr., D.

Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding provedmore » to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.« less

Fission gas bubble identification using MATLAB's image processing toolbox

DOE PAGES

Collette, R.; King, J.; Keiser, Jr., D.; ...

2016-06-08

Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding provedmore » to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.« less

Fission fragment driven neutron source

DOEpatents

Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

1976-01-01

Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

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

Gan, J.; Keiser, D. D.; Miller, B. D.

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

DOE PAGES

Gan, J.; Keiser, D. D.; Miller, B. D.; ...

2017-07-15

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

NASA Astrophysics Data System (ADS)

Carter, Lukas M.

Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or profile measurements of diffusion coefficients.

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

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, V.

1984-06-13

This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

Pyrometer mount for a closed-circuit thermal medium cooled gas turbine

DOEpatents

Jones, Raymond Joseph; Kirkpatrick, Francis Lawrence; Burns, James Lee; Fulton, John Robert

2002-01-01

A steam-cooled second-stage nozzle segment has an outer band and an outer cover defining a plenum therebetween for receiving cooling steam for flow through the nozzles to the inner band and cover therefor and return flow through the nozzles. To measure the temperature of the buckets of the stage forwardly of the nozzle stage, a pyrometer boss is electron beam-welded in an opening through the outer band and TIG-welded to the outer cover plate. By machining a hole through the boss and seating a linearly extending tube in the boss, a line of sight between a pyrometer mounted on the turbine frame and the buckets is provided whereby the temperature of the buckets can be ascertained. The welding of the boss to the outer band and outer cover enables steam flow through the plenum without leakage, while providing a line of sight through the outer cover and outer band to measure bucket temperature.

Aerosol distribution apparatus

DOEpatents

Hanson, W.D.

An apparatus for uniformly distributing an aerosol to a plurality of filters mounted in a plenum, wherein the aerosol and air are forced through a manifold system by means of a jet pump and released into the plenum through orifices in the manifold. The apparatus allows for the simultaneous aerosol-testing of all the filters in the plenum.

Focus control system for stretched-membrane mirror module

DOEpatents

Butler, B.L.; Beninga, K.J.

1991-05-21

A focus control system dynamically sets and controls the focal length of a reflective membrane supported between a perimeter frame. A rear membrane is also supported between the perimeter frame rearward and spaced apart from a back side of the reflective membrane. The space between the membranes defines a plenum space into which a mass of gas at a first pressure is inserted. The pressure differential between the first pressure and an external pressure, such as the atmospheric pressure, causes the reflective membrane to assume a first curvature relative to a reference plane associated with the perimeter frame. This curvature defines the focal length of the reflective membrane. The focal length is dynamically controlled by changing the volume of the plenum space, thereby changing the first pressure. The system can be used to change or maintain the pressure differential and hence the front membrane curvature. The plenum volume is changed by pushing or pulling on a central section of the rear membrane using a suitable actuator. Sensing means continuously sense the location of the reflective membrane relative to the reference plane. This sensed position is compared to a reference position, and a resulting error signal, comprising the difference between the sensed position and reference position, drives the actuator in a direction to minimize the difference. A vent value compensates for temperature changes or leaks in the closed volume by allowing the pressure differential to be adjusted as required to center the working range of the actuator about the desired focal length. 13 figures.

Focus control system for stretched-membrane mirror module

DOEpatents

Butler, Barry L.; Beninga, Kelly J.

1991-01-01

A focus control system dynamically sets and controls the focal length of a reflective membrane supported between a perimeter frame. A rear membrane is also supported between the perimeter frame rearward and spaced apart from a back side of the reflective membrane. The space between the membranes defines a plenum space into which a mass of gas at a first pressure is inserted. The pressure differential between the first pressure and an external pressure, such as the atmospheric pressure, causes the reflective membrane to assume a first curvature relative to a reference plane associated with the perimeter frame. This curvature defines the focal length of the reflective membrane. The focal length is dynamically controlled by changing the volume of the plenum space, thereby changing the first pressure. The system can be used to change or maintain the pressure differential and hence the front membrane curvature. The plenum volume is changed by pushing or pulling on a central section of the rear membrane using a suitable actuator. Sensing means continuously sense the location of the reflective membrane relative to the reference plane. This sensed position is compared to a reference position, and a resulting error signal, comprising the difference between the sensed position and reference position, drives the actuator in a direction to minimize the difference. A vent value compensates for temperature changes or leaks in the closed volume by allowing the pressure differential to be adjusted as required to center the working range of the actuator about the desired focal length.

Light-water-reactor safety research program. Quarterly progress report, July--September 1975

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

Not Available

1975-01-01

Progress is summarized in the following research and development areas: (1) loss-of-coolant accident research; heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; and (3) mechanical properties of Zircaloy containing oxygen. Also included is an appendix on Kinetics of Fission Gas and Volatile Fission-product Behavior under Transient Conditions in LWR Fuel.

Dual-phase reactor plant with partitioned isolation condenser

DOEpatents

Hui, Marvin M.

1992-01-01

A nuclear energy plant housing a boiling-water reactor utilizes an isolation condenser in which a single chamber is partitioned into a distributor plenum and a collector plenum. Steam accumulates in the distributor plenum and is conveyed to the collector plenum through an annular manifold that includes tubes extending through a condenser pool. The tubes provide for a transfer of heat from the steam, forming a condensate. The chamber has a disk-shaped base, a cylindrical sidewall, and a semispherical top. This geometry results in a compact design that exhibits significant performance and cost advantages over prior designs.

High-Flow Jet Exit Rig Designed and Fabricated

NASA Technical Reports Server (NTRS)

Buehrle, Robert J.; Trimarchi, Paul A.

2003-01-01

The High-Flow Jet Exit Rig at the NASA Glenn Research Center is designed to test single flow jet nozzles and to measure the appropriate thrust and noise levels. The rig has been designed for the maximum hot condition of 16 lbm/sec of combustion air at 1960 R (maximum) and to produce a maximum thrust of 2000 lb. It was designed for cold flow of 29.1 lbm/sec of air at 530 R. In addition, it can test dual-flow nozzles (nozzles with bypass flow in addition to core flow) with independent control of each flow. The High- Flow Jet Exit Rig was successfully fabricated in late 2001 and is being readied for checkout tests. The rig will be installed in Glenn's Aeroacoustic Propulsion Laboratory. The High-Flow Jet Exit Rig consists of the following major components: a single component force balance, the natural-gas-fueled J-79 combustor assembly, the plenum and manifold assembly, an acoustic/instrumentation/seeding (A/I/S) section, a table, and the research nozzles. The rig will be unique in that it is designed to operate uncooled. The structure survives the 1960 R test condition because it uses carefully selected high temperature alloy materials such as Hastelloy-X. The lower plenum assembly was designed to operate at pressures to 450 psig at 1960 R, in accordance with the ASME B31.3 piping code. The natural gas-fueled combustor fires directly into the lower manifold. The hot air is directed through eight 1-1/2-in. supply pipes that supply the upper plenum. The flow is conditioned in the upper plenum prior to flowing to the research nozzle. The 1-1/2-in. supply lines are arranged in a U-shaped design to provide for a flexible piping system. The combustor assembly checkout was successfully conducted in Glenn's Engine Component Research Laboratory in the spring of 2001. The combustor is a low-smoke version of the J79 combustor used to power the F4 Phantom military aircraft. The natural gas-fueled combustor demonstrated high-efficiency combustion over a wide range of operating conditions. This wide operating envelope is required to support the testing of both single- and dual-flow nozzles. Key research goals include providing simultaneous, highly accurate acoustic, flow, and thrust measurements on jet nozzle models in realistic flight conditions, as well as providing scaleable acoustic results. The High-Flow Jet Exit Rig is a second-generation high-flow test rig. Improvements include cleaner flow with reduced levels of particulate, soot, and odor. Choked-flow metering is required with plus or minus 0.25-percent accuracy. Thrust measurements from 0 to 2000 lbf are required with plus or minus 0.25-percent accuracy. Improved acoustics will be achieved by minimizing noise through large pipe bend radii, lower internal flow velocities, and microdrilled choke plates with thousands of 0.040-in.- diameter holes.

Vacuum testing of a miniaturised switch mode amplifier powering an electrothermal plasma micro-thruster

NASA Astrophysics Data System (ADS)

Charles, Christine; Liang, Wei; Raymond, Luke; Rivas-Davila, Juan; Boswell, Roderick W.

2017-08-01

A structurally supportive miniaturised low-weight (≤150 g) radiofrequency switch mode amplifier developed to power the small diameter Pocket Rocket electrothermal plasma micro-thruster called MiniPR is tested in vacuum conditions representative of space to demonstrate its suitability for use on nano-satellites such as `CubeSats'. Argon plasma characterisation is carried out by measuring the optical emission signal seen through the plenum window versus frequency (12.8-13.8 MHz) and the plenum cavity pressure increase (indicative of thrust generation from volumetric gas heating in the plasma cavity) versus power (1-15 Watts) with the amplifier operating at atmospheric pressure and a constant flow rate of 20 sccm. Vacuum testing is subsequently performed by measuring the operational frequency range of the amplifier as a function of gas flow rate. The switch mode amplifier design is finely tuned to the input impedance of the thruster ˜16 pF) to provide a power efficiency of 88 % at the resonant frequency and a direct feed to a low-loss (˜ 10 %) impedance matching network. This system provides successful plasma coupling at 1.54 Watts for all investigated flow rates (10-130 sccm) for cryogenic pumping speeds of the order of 6000 l.s^{-1} and a vacuum pressure of the order of ˜ 2x10^{-5} Torr during operation. Interestingly, the frequency bandwidth for which a plasma can be coupled increases from 0.04 to 0.4 MHz when the gas flow rate is increased, probably as a result of changes in the plasma impedance.

A compact gas-filled avalanche counter for DANCE

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

Wu, C. Y.; Chyzh, A.; Kwan, E.

2012-08-04

A compact gas-filled avalanche counter for the detection of fission fragments was developed for a highly segmented 4π γ-ray calorimeter, namely the Detector for Advanced Neutron Capture Experiments located at the Lujan Center of the Los Alamos Neutron Science Center. It has been used successfully for experiments with 235U, 238Pu, 239Pu, and 241Pu isotopes to provide a unique signature to differentiate the fission from the competing neutron-capture reaction channel. We also used it to study the spontaneous fission in 252Cf. The design and performance of this avalanche counter for targets with extreme α-decay rate up to ~2.4×108/s are described.

Application of the DART Code for the Assessment of Advanced Fuel Behavior

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

Rest, J.; Totev, T.

2007-07-01

The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO{sub 2}more » fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)« less

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

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

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety ofmore » LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.« less

Cold plate

DOEpatents

Marroquin, Christopher M.; O'Connell, Kevin M.; Schultz, Mark D.; Tian, Shurong

2018-02-13

A cold plate, an electronic assembly including a cold plate, and a method for forming a cold plate are provided. The cold plate includes an interface plate and an opposing plate that form a plenum. The cold plate includes a plurality of active areas arranged for alignment over respective heat generating portions of an electronic assembly, and non-active areas between the active areas. A cooling fluid flows through the plenum. The plenum, at the non-active areas, has a reduced width and/or reduced height relative to the plenum at the active areas. The reduced width and/or height of the plenum, and exterior dimensions of cold plate, at the non-active areas allow the non-active areas to flex to accommodate surface variations of the electronics assembly. The reduced width and/or height non-active areas can be specifically shaped to fit between physical features of the electronics assembly.

Combustor assembly for use in a turbine engine and methods of assembling same

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2013-05-14

A fuel nozzle assembly for use with a turbine engine is described herein. The fuel nozzle assembly includes a plurality of fuel nozzles positioned within an air plenum defined by a casing. Each of the plurality of fuel nozzles is coupled to a combustion liner defining a combustion chamber. Each of the plurality of fuel nozzles includes a housing that includes an inner surface that defines a cooling fluid plenum and a fuel plenum therein, and a plurality of mixing tubes extending through the housing. Each of the mixing tubes includes an inner surface defining a flow channel extending between the air plenum and the combustion chamber. At least one mixing tube of the plurality of mixing tubes including at least one cooling fluid aperture for channeling a flow of cooling fluid from the cooling fluid plenum to the flow channel.

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

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

DOE PAGES

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian; ...

2017-02-27

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

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

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Multiscale simulation of xenon diffusion and grain boundary segregation in UO₂

DOE PAGES

Andersson, David A.; Tonks, Michael R.; Casillas, Luis; ...

2015-07-01

In light water reactor fuel, gaseous fission products segregate to grain boundaries, resulting in the nucleation and growth of large intergranular fission gas bubbles. The segregation rate is controlled by diffusion of fission gas atoms through the grains and interaction with the boundaries. Based on the mechanisms established from earlier density functional theory (DFT) and empirical potential calculations, diffusion models for xenon (Xe), uranium (U) vacancies and U interstitials in UO₂ have been derived for both intrinsic (no irradiation) and irradiation conditions. Segregation of Xe to grain boundaries is described by combining the bulk diffusion model with a model formore » the interaction between Xe atoms and three different grain boundaries in UO₂ (Σ5 tilt, Σ5 twist and a high angle random boundary), as derived from atomistic calculations. The present model does not attempt to capture nucleation or growth of fission gas bubbles at the grain boundaries. The point defect and Xe diffusion and segregation models are implemented in the MARMOT phase field code, which is used to calculate effective Xe and U diffusivities as well as to simulate Xe redistribution for a few simple microstructures.« less

TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

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

Jian Gan; Brandon Miller; Dennis Keiser

2014-04-01

As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists ofmore » fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.« less

Quantity of 135I released from the AGR-1, AGR-2, and AGR-3/4 experiments and discovery of 131I at the FPMS traps during the AGR-3/4 experiment

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

Scates, Dawn M.

2014-09-01

A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission productmore » release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay with the parent half life. This equilibrium condition enables the determination of the amount of 135I released from the fuel particles by measurement of the 135mXe at the FPM following reactor shutdown. In this paper, the 135I released will be reported and compared to similar releases for noble gases as well as the unexpected finding of 131I deposition from intentional impure gas injection into capsule 11 of experiment AGR 3/4.« less

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

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

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

Navier-Stokes simulation of real gas flows in nozzles

NASA Technical Reports Server (NTRS)

Nagaraj, N.; Lombard, C. K.

1987-01-01

Air flow in a hypersonic nozzle causes real gas effects due to reaction among the species constituting air. Such reactions may be in chemical equilibrium or in chemical nonequilibrium. Here using the CSCM upwind scheme for the compressible Navier-Stokes equations, the real gas flowfield in an arcjet nozzle is computed for both the equilibrium case and the nonequilibrium case. A hypersonic nozzle flow arising from a pebble bed heated plenum is also computed for the equilibrium situation. Between the equilibrium cases, the chemistry is treated by two different schemes and comments are made as to computational complexity. For the nonequilibrium case, a full set of seventeen reactions and full implicit coupling of five species with gasdynamics is employed to compute the flowfield. For all cases considered here the gas is assumed to be a calorically imperfect mixture of ideal gases in thermal equilibrium.

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.

Effects of Hole Length, Supply Plenum Geometry, and Freestream Turbulence on Film Cooling Performance

NASA Technical Reports Server (NTRS)

Burd, Steven W.; Simon, Terrence W.; Thurman, Douglas (Technical Monitor)

2000-01-01

Experimental measurements are presented in this report to document the sensitivity of film cooling performance to the hole length and coolant delivery plenum geometry. Measurements with hot-wire anemometry detail velocity, local turbulence, and spectral distributions over the exit plane of film cooling holes and downstream of injection in the coolant-freestream interaction zone. Measurements of discharge coefficients and adiabatic effectiveness are also provided. Coolant is supplied to the film cooling holes by means of a large, open plenum and through plenums which force the coolant to approach the holes either co-current or counter-current to the freestream. A single row of film cooling holes with 35 degree-inclined streamwise at two coolant-to-freestream velocity ratios, 0.5 and 1.0, is investigated. The coolant-to-freestream density ratio is maintained in the range 0.96 to 1.0. Measurements were taken under high-freestream (FSTI = 12%) and low-freestream turbulence intensity (FSTI = 0.5%) conditions. The results document the effects of the hole L/D, coolant supply plenum geometry, velocity ratio, and FSTI. In general, hole L/D and the supply plenum geometry play influential roles in the film cooling performance. Hole L/D effects, however, are more pronounced. Film cooling performance is also dependent upon the velocity ratio and FSTI.

Research on fission fragment excitation of gases and nuclear pumping of lasers

NASA Technical Reports Server (NTRS)

Schneider, R. T.; Davie, R. N.; Davis, J. F.; Fuller, J. L.; Paternoster, R. R.; Shipman, G. R.; Sterritt, D. E.; Helmick, H. H.

1974-01-01

Experimental investigations of fission fragment excited gases are reported along with a theoretical analysis of population inversions in fission fragment excited helium. Other studies reported include: nuclear augmentation of gas lasers, direct nuclear pumping of a helium-xenon laser, measurements of a repetitively pulsed high-power CO2 laser, thermodynamic properties of UF6 and UF6/He mixtures, and nuclear waste disposal utilizing a gaseous core reactor.

Production of Sn and Sb isotopes in high-energy neutron-induced fission of natU

NASA Astrophysics Data System (ADS)

Mattera, A.; Pomp, S.; Lantz, M.; Rakopoulos, V.; Solders, A.; Al-Adili, A.; Penttilä, H.; Moore, I. D.; Rinta-Antila, S.; Eronen, T.; Kankainen, A.; Pohjalainen, I.; Gorelov, D.; Canete, L.; Nesterenko, D.; Vilén, M.; Äystö, J.

2018-03-01

The first systematic measurement of neutron-induced fission yields has been performed at the upgraded IGISOL-4 facility at the University of Jyväskylä, Finland. The fission products from high-energy neutron-induced fission of nat U were stopped in a gas cell filled with helium buffer gas, and were online separated with a dipole magnet. The isobars, with masses in the range A = 128-133 , were transported to a tape-implantation station and identified using γ -spectroscopy. We report here the relative cumulative isotopic yields of tin ( Z = 50) and the relative independent isotopic yields of antimony ( Z = 51) . Isomeric yield ratios were also obtained for five nuclides. The yields of tin show a staggered behaviour around A = 131 , not observed in the ENDF/B-VII.1 evaluation. The yields of antimony also contradict the trend from the evaluation, but are in agreement with a calculation performed using the GEF model that shows the yield increasing with mass in the range A = 128-133.

Development of a Gas Filled Magnet spectrometer coupled with the Lohengrin spectrometer for fission study

NASA Astrophysics Data System (ADS)

Kessedjian, G.; Chebboubi, A.; Faust, H.; Köster, U.; Materna, T.; Sage, C.; Serot, O.

2013-03-01

The accurate knowledge of the fission of actinides is necessary for studies of innovative nuclear reactor concepts. The fission yields have a direct influence on the evaluation of the fuel inventory or the reactor residual power after shutdown. A collaboration between the ILL, LPSC and CEA has developed a measurement program on fission fragment distributions at ILL in order to measure the isotopic and isomeric yields. The method is illustrated using the 233U(n,f)98Y reaction. However, the extracted beam from the Lohengrin spectrometer is not isobaric ions which limits the low yield measurements. Presently, the coupling of the Lohengrin spectrometer with a Gas Filled Magnet (GFM) is studied at the ILL in order to define and validate the enhanced purification of the extracted beam. This work will present the results of the spectrometer characterisation, along with a comparison with a dedicated Monte Carlo simulation especially developed for this purpose.

Simulation of radiation driven fission gas diffusion in UO 2, ThO 2 and PuO 2

DOE PAGES

Cooper, Michael William D.; Stanek, Christopher Richard; Turnbull, James Anthony; ...

2016-12-01

Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. Here we present a molecular dynamics (MD) study of Xe, Kr, Th, U, Pu and O diffusion due to irradiation. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Thermal spike simulations are used to confirm that electronic stopping remedies the discrepancy with experiment and the predicted diffusivities lie within the scatter of the experimental data. Here, our results predict that the diffusion coefficients are ordered such that D* 0more » > D* Kr > D* Xe > D* U. For all species >98.5% of diffusivity is accounted for by electronic stopping. Fission gas diffusivity was not predicted to vary significantly between ThO 2, UO 2 and PuO 2, indicating that this process would not change greatly for mixed oxide fuels.« less

Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.

2012-06-01

Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

Transient Effects in Turbulence Modelling.

DTIC Science & Technology

1979-12-01

plenum region of a liquid-metal- cooled fast breeder reactor (LMFBR). The efficient heat transfer characteristics of liquid metal coolant, combined...Transients in Generalized Liquid-Metal Fast Breeder Reactor Outlet Plenums," Nuclear Technology, Vol. 44, July 1979, p. 210. 135 15. Lorenz, J. J., "MIX... Sodium Coolant in the Outlet Plenum of a Fast Nuclear Reactor ," Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 1565-1579. 19. Chen, Y. B., Golay, M. W

Film cooling from inclined cylindrical holes using large eddy simulations

NASA Astrophysics Data System (ADS)

Peet, Yulia V.

2006-12-01

The goal of the present study is to investigate numerically the physics of the flow, which occurs during the film cooling from inclined cylindrical holes, Film cooling is a technique used in gas turbine industry to reduce heat fluxes to the turbine blade surface. Large Eddy Simulation (LES) is performed modeling a realistic film cooling configuration, which consists of a large stagnation-type reservoir, feeding an array of discrete cooling holes (film holes) flowing into a flat plate turbulent boundary layer. Special computational methodology is developed for this problem, involving coupled simulations using multiple computational codes. A fully compressible LES code is used in the area above the flat plate, while a low Mach number LES code is employed in the plenum and film holes. The motivation for using different codes comes from the essential difference in the nature of the flow in these different regions. Flowfield is analyzed inside the plenum, film hole and a crossflow region. Flow inside the plenum is stagnating, except for the region close to the exit, where it accelerates rapidly to turn into the hole. The sharp radius of turning at the trailing edge of the plenum pipe connection causes the flow to separate from the downstream wall of the film hole. After coolant injection occurs, a complex flowfield is formed consisting of coherent vortical structures responsible for bringing hot crossflow fluid in contact with the walls of either the film hole or the blade, thus reducing cooling protection. Mean velocity and turbulent statistics are compared to experimental measurements, yielding good agreement for the mean flowfield and satisfactory agreement for the turbulence quantities. LES results are used to assess the applicability of basic assumptions of conventional eddy viscosity turbulence models used with Reynolds-averaged (RANS) approach, namely the isotropy of an eddy viscosity and thermal diffusivity. It is shown here that these assumptions do not hold for the film cooling flows. Comparison of film cooling effectiveness with experiments shows fair agreement for the centerline and laterally-averaged effectiveness. Lateral growth of the jet as judged from the lateral distribution of effectiveness is predicted correctly.

Open cycle gas core nuclear rockets

NASA Technical Reports Server (NTRS)

Ragsdale, Robert

1991-01-01

The open cycle gas core engine is a nuclear propulsion device. Propulsion is provided by hot hydrogen which is heated directly by thermal radiation from the nuclear fuel. Critical mass is sustained in the uranium plasma in the center. It has typically 30 to 50 kg of fuel. It is a thermal reactor in the sense that fissions are caused by absorption of thermal neutrons. The fast neutrons go out to an external moderator/reflector material and, by collision, slow down to thermal energy levels, and then come back in and cause fission. The hydrogen propellant is stored in a tank. The advantage of the concept is very high specific impulse because you can take the plasma to any temperature desired by increasing the fission level by withdrawing or turning control rods or control drums.

Irradiation behaviour of the large grained UO{sub 2} fuel pellet in the transient conditions

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

Kosaka, Yuji; Watanabe, Seiichi; Arakawa, Yasushi

2007-07-01

In order to achieve a high duty fuel rod design, it is the key issue to suppress the fission gas release from the view point of the fuel rod inner pressure design. The large grain UO{sub 2} pellet is one of the candidates to meet such a requirement by reducing the fission gas release especially at high power and/or high burnup. We have demonstrated the fuel performance of the large grain pellet in the PWR irradiation conditions, which was fabricated with no additive but with active UO{sub 2} powder through the conventional pelletizing process for the normal grain size pellet.more » According to the mechanism of the fission gas retention, there may be a concern about the larger gas bubble swelling of the large grain pellet at the power transient conditions which may increase the potential of the PCMI failure. In this paper, we focus on the differences of the dimensional change in comparison among the pellets with the different grain sizes at the power transient conditions. The power ramp tests were carried out on the high burnup fuel rods of normal and large grain pellet with no additive, which had been irradiated in the PWR conditions up to around 60 GWd/t at peak position. The detailed PIE results revealed that the volume increment due to the power ramp clearly showed the dependence on the grain size as well as the fission gas release and suggested that the larger grain with no additive may suppress the gas bubble swelling at the power transient conditions. According to the experimental results, it is concluded that the large grain pellet with no additive does not deteriorate the irradiation performance during the power transient conditions from the view point of the gas bubble swelling. (authors)« less

Stability Study of the RERTR Fuel Microstructure

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

Jian Gan; Dennis Keiser; Brandon Miller

2014-04-01

The irradiation stability of the interaction phases at the interface of fuel and Al alloy matrix as well as the stability of the fission gas bubble superlattice is believed to be very important to the U-Mo fuel performance. In this paper the recent result from TEM characterization of Kr ion irradiated U-10Mo-5Zr alloy will be discussed. The focus will be on the phase stability of Mo2-Zr, a dominated second phase developed at the interface of U-10Mo and the Zr barrier in a monolithic fuel plate from fuel fabrication. The Kr ion irradiations were conducted at a temperature of 200 degreesmore » C to an ion fluence of 2.0E+16 ions/cm2. To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated dispersion U-7Mo fuel and monolithic U-10Mo fuel, a FIB-TEM sample of the irradiated U-10Mo fuel (3.53E+21 fission/cm3) was used for a TEM in-situ heating experiment. The preliminary result showed extraordinary thermal stability of the fission gas bubble superlattice. The implication of the TEM observation from these two experiments on the fuel microstructural evolution under irradiation will be discussed.« less

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

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

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

Paint-Overspray Catcher

NASA Technical Reports Server (NTRS)

Weinstein, Leonard M.

2003-01-01

An apparatus to catch paint overspray has been proposed. Overspray is an unavoidable parasitic component of spray that occurs because the flow of air or other gas in the spray must turn at the sprayed surface. Very small droplets are carried away in this turning flow, and some land on adjacent surfaces not meant to be painted. The basic principle of the paint-spray catcher is to divert the overspray into a suction system at the boundary of the area to be painted. The paint-spray catcher (see figure) would include a toroidal plenum connected through narrow throat to a nozzle that would face toward the center of the torus, which would be positioned over the center of the area to be spray-painted. The plenum would be supported by four tubes that would also serve as suction exhaust ducts. The downstream ends of the tubes (not shown in the figure) would be connected to a filter on a suction pump. The pump would be rated to provide a suction mass flow somewhat greater than that of the directed spray gas stream, so that the nozzle would take in a small excess of surrounding gas and catch nearly all of the overspray. A small raised lip at the bottom edge of the nozzle would catch paint that landed inside the nozzle. Even if the paint is directly piston pumped, the droplets entrain an air flow by time they approach the wall, so there is always a gas stream to carry the excess droplets to the side. For long-duration spraying operations, it could be desirable to include a suction-drain apparatus to prevent overflowing and dripping of paint from inside the lip. A version without an external contraction and with the throat angled downward would be a more compact version of catcher, although it might be slightly less efficient.

3D CFD ELECTROCHEMICAL AND HEAT TRANSFER MODEL OF AN INTERNALLY MANIFOLDED SOLID OXIDE ELECTROLYSIS CELL

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

Grant L. Hawkes; James E. O'Brien; Greg Tao

2011-11-01

A three-dimensional computational fluid dynamics (CFD) electrochemical model has been created to model high-temperature electrolysis cell performance and steam electrolysis in an internally manifolded planar solid oxide electrolysis cell (SOEC) stack. This design is being evaluated at the Idaho National Laboratory for hydrogen production from nuclear power and process heat. Mass, momentum, energy, and species conservation and transport are provided via the core features of the commercial CFD code FLUENT. A solid-oxide fuel cell (SOFC) model adds the electrochemical reactions and loss mechanisms and computation of the electric field throughout the cell. The FLUENT SOFC user-defined subroutine was modified formore » this work to allow for operation in the SOEC mode. Model results provide detailed profiles of temperature, operating potential, steam-electrode gas composition, oxygen-electrode gas composition, current density and hydrogen production over a range of stack operating conditions. Single-cell and five-cell results will be presented. Flow distribution through both models is discussed. Flow enters from the bottom, distributes through the inlet plenum, flows across the cells, gathers in the outlet plenum and flows downward making an upside-down ''U'' shaped flow pattern. Flow and concentration variations exist downstream of the inlet holes. Predicted mean outlet hydrogen and steam concentrations vary linearly with current density, as expected. Effects of variations in operating temperature, gas flow rate, oxygen-electrode and steam-electrode current density, and contact resistance from the base case are presented. Contour plots of local electrolyte temperature, current density, and Nernst potential indicate the effects of heat transfer, reaction cooling/heating, and change in local gas composition. Results are discussed for using this design in the electrolysis mode. Discussion of thermal neutral voltage, enthalpy of reaction, hydrogen production, cell thermal efficiency, cell electrical efficiency, and Gibbs free energy are discussed and reported herein.« less

Fission in the landscape of heaviest elements: Some recent examples

NASA Astrophysics Data System (ADS)

Khuyagbaatar, J.; Yakushev, A.; Düllmann, Ch. E.; Ackermann, D.; Andersson, L.-L.; Block, M.; Brand, H.; Even, J.; Forsberg, U.; Hartmann, W.; Herzberg, R.-D.; Heßberger, F. P.; Hoffmann, J.; Hübner, A.; Jäger, E.; Jeppsson, J.; Kindler, B.; Kratz, J. V.; Krier, J.; Kurz, N.; Lommel, B.; Maiti, M.; Minami, S.; Rudolph, D.; Runke, J.; Sarmiento, L. G.; Schädel, M.; Schausten, B.; Steiner, J.; Heidenreich, T. Torres De; Uusitalo, J.; Wiehl, N.; Yakusheva, V.

2016-12-01

The fission process still remains a main factor that determines the stability of the atomic nucleus of heaviest elements. Fission half-lives vary over a wide range, 10-19-1024 s. Present experimental techniques for the synthesis of the superheavy elements that usually measure α-decay chains are sensitive only in a limited range of half-lives, often 10-5-103 s. In the past years, measurement techniques for very short-lived and very long-lived nuclei were significantly improved at the gas-filled recoil separator TASCA at GSI Darmstadt. Recently, several experimental studies of fission-related phenomena have successfully been performed. In this paper, results on 254-256Rf and 266Lr are presented and corresponding factors for retarding the fission process are discussed.

Cascading Tesla Oscillating Flow Diode for Stirling Engine Gas Bearings

NASA Technical Reports Server (NTRS)

Dyson, Rodger

2012-01-01

Replacing the mechanical check-valve in a Stirling engine with a micromachined, non-moving-part flow diode eliminates moving parts and reduces the risk of microparticle clogging. At very small scales, helium gas has sufficient mass momentum that it can act as a flow controller in a similar way as a transistor can redirect electrical signals with a smaller bias signal. The innovation here forces helium gas to flow in predominantly one direction by offering a clear, straight-path microchannel in one direction of flow, but then through a sophisticated geometry, the reversed flow is forced through a tortuous path. This redirection is achieved by using microfluid channel flow to force the much larger main flow into this tortuous path. While microdiodes have been developed in the past, this innovation cascades Tesla diodes to create a much higher pressure in the gas bearing supply plenum. In addition, the special shape of the leaves captures loose particles that would otherwise clog the microchannel of the gas bearing pads.

Stagnation region gas film cooling: Spanwise angled injection from multiple rows of holes. [gas turbine engines

NASA Technical Reports Server (NTRS)

Luckey, D. W.; Lecuyer, M. R.

1981-01-01

The stagnation region of a cylinder in a cross flow was used in experiments conducted with both a single row and multiple rows of spanwise angled (25 deg) coolant holes for a range of the coolant blowing ratio with a freestream to wall temperature ratio approximately equal to 1.7 and R(eD) = 90,000. Data from local heat flux measurements are presented for injection from a single row located at 5 deg, 22.9 deg, 40.8 deg, 58.7 deg from stagnation using a hole spacing ratio of S/d(o) = 5 and 10. Three multiple row configurations were also investigated. Data are presented for a uniform blowing distribution and for a nonuniform blowing distribution simulating a plenum supply. The data for local Stanton Number reduction demonstrated a lack of lateral spreading by the coolant jets. Heat flux levels larger than those without film cooling were observed directly behind the coolant holes as the blowing ratio exceeded a particular value. The data were spanwise averaged to illustrate the influence of injection location, blowing ratio and hole spacing. The large values of blowing ratio for the blowing distribution simulating a plenum supply resulted in heat flux levels behind the holes in excess of the values without film cooling. An increase in freestream turbulence intensity from 4.4 to 9.5 percent had a negligible effect on the film cooling performance.

Air lifted and propelled vehicle

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

Jones, T.E.; Johnson, R.A.

1987-02-17

This patent describes a vehicle which rides on air cushion and which is propelled by air, comprising: upper deck means, having a bottom edge which defines the periphery of an area; a thin, flexible sheet located below the upper deck means, extending beneath the bottom edge and secured beneath the bottom edge for defining a plenum that is defined by and closed off by the upper deck means and the sheet. The deck means is shaped within the area defined by its bottom edge for causing the plenum to always be an open space and the upper deck means ismore » rigid enough to maintain that open condition of the plenum; the sheet being secured in a manner permitting the sheet to pillow when air is pressurized in the plenum; and the sheet being perforated below the upper deck means for permitting exit of air from the plenum at a controllable rate through the perforations; the sheet having a large plurality of the perforations dispersed over most of its area below the upper deck means; each of the perforations being a hole.« less

Large Eddy Simulation of a Film Cooling Technique with a Plenum

NASA Astrophysics Data System (ADS)

Dharmarathne, Suranga; Sridhar, Narendran; Araya, Guillermo; Castillo, Luciano; Parameswaran, Sivapathasund

2012-11-01

Factors that affect the film cooling performance have been categorized into three main groups: (i) coolant & mainstream conditions, (ii) hole geometry & configuration, and (iii) airfoil geometry Bogard et al. (2006). The present study focuses on the second group of factors, namely, the modeling of coolant hole and the plenum. It is required to simulate correct physics of the problem to achieve more realistic numerical results. In this regard, modeling of cooling jet hole and the plenum chamber is highly important Iourokina et al. (2006). Substitution of artificial boundary conditions instead of correct plenum design would yield unrealistic results Iourokina et al. (2006). This study attempts to model film cooling technique with a plenum using a Large Eddy Simulation.Incompressible coolant jet ejects to the surface of the plate at an angle of 30° where it meets compressible turbulent boundary layer which simulates the turbine inflow conditions. Dynamic multi-scale approach Araya (2011) is introduced to prescribe turbulent inflow conditions. Simulations are carried out for two different blowing ratios and film cooling effectiveness is calculated for both cases. Results obtained from LES will be compared with experimental results.

Pre-mixing apparatus for a turbine engine

DOEpatents

Lacy, Benjamin Paul [Greer, SC; Varatharajan, Balachandar [Cincinnati, OH; Ziminsky, Willy Steve [Simpsonville, SC; Kraemer, Gilbert Otto [Greer, SC; Yilmaz, Ertan [Albany, NY; Melton, Patrick Benedict [Horse Shoe, NC; Zuo, Baifang [Simpsonville, SC; Stevenson, Christian Xavier [Inman, SC; Felling, David Kenton [Greenville, SC; Uhm, Jong Ho [Simpsonville, SC

2012-04-03

A pre-mixing apparatus for a turbine engine includes a main body having an inlet portion, an outlet portion and an exterior wall that collectively establish at least one fluid delivery plenum, and a plurality of fluid delivery tubes extending through at least a portion of the at least one fluid delivery plenum. Each of the plurality of fluid delivery tubes includes at least one fluid delivery opening fluidly connected to the at least one fluid delivery plenum. With this arrangement, a first fluid is selectively delivered to the at least one fluid delivery plenum, passed through the at least one fluid delivery opening and mixed with a second fluid flowing through the plurality of fluid delivery tubes prior to being combusted in a combustion chamber of a turbine engine.

Fission fragment excited laser system

DOEpatents

McArthur, David A.; Tollefsrud, Philip B.

1976-01-01

A laser system and method for exciting lasing action in a molecular gas lasing medium which includes cooling the lasing medium to a temperature below about 150 K and injecting fission fragments through the lasing medium so as to preferentially excite low lying vibrational levels of the medium and to cause population inversions therein. The cooled gas lasing medium should have a mass areal density of about 5 .times. 10.sup.-.sup.3 grams/square centimeter, relaxation times of greater than 50 microseconds, and a broad range of excitable vibrational levels which are excitable by molecular collisions.

Thermodynamic properties of a high pressure subcritical UF6/He gas volume (irradiated by an external source)

NASA Technical Reports Server (NTRS)

Sterritt, D. E.; Lalos, G. T.; Schneider, R. T.

1976-01-01

A computer simulation study concerning a compressed fissioning UF6 gas is presented. The compression is to be achieved by a ballistic piston compressor. Data on UF6 obtained with this compressor were incorporated in the simulation study. As a neutron source to create the fission events in the compressed gas, a fast burst reactor was considered. The conclusion is that it takes a neutron flux in excess of 10 to the 15th power n/sec sq cm to produce measurable increases in pressure and temperature, while a flux in excess of 10 to 19th power n/sq cm sec would probably damage the compressor.

Cryogenic system with GM cryocooler for krypton, xenon separation from hydrogen-helium purge gas

NASA Astrophysics Data System (ADS)

Chu, X. X.; Zhang, M. M.; Zhang, D. X.; Xu, D.; Qian, Y.; Liu, W.

2014-01-01

In the thorium molten salt reactor (TMSR), fission products such as krypton, xenon and tritium will be produced continuously in the process of nuclear fission reaction. A cryogenic system with a two stage GM cryocooler was designed to separate Kr, Xe, and H2 from helium purge gas. The temperatures of two stage heat exchanger condensation tanks were maintained at about 38 K and 4.5 K, respectively. The main fluid parameters of heat transfer were confirmed, and the structural heat exchanger equipment and cold box were designed. Designed concentrations after cryogenic separation of Kr, Xe and H2 in helium recycle gas are less than 1 ppb.

Thermodynamic properties of a high pressure subcritical UF/sub 6/He gas volume (irradiated by an external source)

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

Sterritt, D.E.; Lalos, G.T.; Schneider, R.T.

1976-12-01

A computer simulation study concerning a compressed fissioning UF/sub 6/ gas is presented. The compression is to be achieved by a ballistic piston compressor. Data on UF/sub 6/ obtained with this compressor were incorporated in the simulation study. As a neutron source to create the fission events in the compressed gas, a fast burst reactor was considered. The conclusion is that it takes a neutron flux in excess of 10/sup 15/ n/cm/sup 2/-s to produce measurable increases in pressure and temperature, while a flux in excess of 10/sup 19/ n/cm/sup 2/-s would probably damage the compressor.

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

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Tuned intake air system for a rotary engine

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

Corbett, W.D.

This patent describes a rotary internal combustion engine for an outboard board motor. It comprises a plenum chamber attached to the rear of the engine; and the plenum chamber including an inner wall attached to the exhaust manifold; an inlet conduit connecting the cooling air exit passage and the inlet air opening; an outlet conduit connecting the outlet air opening and the combustion air inlet; and the outlet conduit terminating in a combustion air outlet in the inner wall of the plenum chamber.

System and method for the analysis of one or more compounds and/or species produced by a solution-based nuclear reactor

DOEpatents

Policke, Timothy A; Nygaard, Eric T

2014-05-06

The present invention relates generally to both a system and method for determining the composition of an off-gas from a solution nuclear reactor (e.g., an Aqueous Homogeneous Reactor (AHR)) and the composition of the fissioning solution from those measurements. In one embodiment, the present invention utilizes at least one quadrupole mass spectrometer (QMS) in a system and/or method designed to determine at least one or more of: (i) the rate of production of at least one gas and/or gas species from a nuclear reactor; (ii) the effect on pH by one or more nitrogen species; (iii) the rate of production of one or more fission gases; and/or (iv) the effect on pH of at least one gas and/or gas species other than one or more nitrogen species from a nuclear reactor.

Production of fissioning uranium plasma to approximate gas-core reactor conditions

NASA Technical Reports Server (NTRS)

Lee, J. H.; Mcfarland, D. R.; Hohl, F.; Kim, K. H.

1974-01-01

The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, Vladimir

1986-01-01

This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

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

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Fission track astrology of three Apollo 14 gas-rich breccias

NASA Technical Reports Server (NTRS)

Graf, H.; Shirck, J.; Sun, S.; Walker, R.

1973-01-01

The three Apollo 14 breccias 14301, 14313, and 14318 all show fission xenon due to the decay of Pu-244. To investigate possible in situ production of the fission gas, an analysis was made of the U-distribution in these three breccias. The major amount of the U lies in glass clasts and in matrix material and no more than 25% occurs in distinct high-U minerals. The U-distribution of each breccia is discussed in detail. Whitlockite grains in breccias 14301 and 14318 found with the U-mapping were etched and analyzed for fission tracks. The excess track densities are much smaller than indicated by the Xe-excess. Because of a preirradiation history documented by very high track densities in feldspar grains, however, it is impossible to attribute the excess tracks to the decay of Pu-244. A modified track method has been developed for measuring average U-concentrations in samples containing a heterogeneous distribution of U in the form of small high-U minerals. The method is briefly discussed, and results for the rocks 14301, 14313, 14318, 68815, 15595, and the soil 64421 are given.

A Gas Turbine Compressor Simulation Model for Inclusion of Active Control Strategies

DTIC Science & Technology

2001-06-01

family is based on a classic ,. ( feedback control system, i.e., a sensor /actuator pair and a suitable control law. Ffowcs Williams and Huang (1989...stabilized the 0,4 unstable flow in a small compression system by using a pressure 0 0.2 04 0.6 0 8 :, .4 16 . a) sensor located in the plenum and a...requirements: very fast response to frequency and a large number of sensors and actuators. The former conflicts with the use of analogic or AA = Kdp/dt

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

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

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

2. PADDLE FAN IN PLENUM INTERIOR. Hot Springs National ...

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

2. PADDLE FAN IN PLENUM INTERIOR. - Hot Springs National Park Bathhouse Row, Maurice Bathhouse: Mechanical & Piping Systems, State Highway 7, 1 mile north of U.S. Highway 70, Hot Springs, Garland County, AR

Fission yield measurements at IGISOL

NASA Astrophysics Data System (ADS)

Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

2016-06-01

The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, Douglas M.; Nesbitt, Loyd B.

1997-01-01

A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

Methods of separating particulate residue streams

DOEpatents

Hoskinson, Reed L [Rigby, ID; Kenney, Kevin L [Idaho Falls, ID; Wright, Christopher T [Idaho Falls, ID; Hess, J Richard [Idaho Falls, ID

2011-04-05

A particulate residue separator and a method for separating a particulate residue stream may include an air plenum borne by a harvesting device, and have a first, intake end and a second, exhaust end; first and second particulate residue air streams that are formed by the harvesting device and that travel, at least in part, along the air plenum and in a direction of the second, exhaust end; and a baffle assembly that is located in partially occluding relation relative to the air plenum and that substantially separates the first and second particulate residue air streams.

Particulate residue separators for harvesting devices

DOEpatents

Hoskinson, Reed L.; Kenney, Kevin L.; Wright, Christopher T.; Hess, John R.

2010-06-29

A particulate residue separator and a method for separating a particulate residue stream may include a plenum borne by a harvesting device, and have a first, intake end and a second, exhaust end; first and second particulate residue air streams which are formed by the harvesting device and which travel, at least in part, along the plenum and in a direction of the second, exhaust end; and a baffle assembly which is located in partially occluding relation relative to the plenum, and which substantially separates the first and second particulate residue air streams.

Cryogenic system with GM cryocooler for krypton, xenon separation from hydrogen-helium purge gas

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

Chu, X. X.; Zhang, D. X.; Qian, Y.

2014-01-29

In the thorium molten salt reactor (TMSR), fission products such as krypton, xenon and tritium will be produced continuously in the process of nuclear fission reaction. A cryogenic system with a two stage GM cryocooler was designed to separate Kr, Xe, and H{sub 2} from helium purge gas. The temperatures of two stage heat exchanger condensation tanks were maintained at about 38 K and 4.5 K, respectively. The main fluid parameters of heat transfer were confirmed, and the structural heat exchanger equipment and cold box were designed. Designed concentrations after cryogenic separation of Kr, Xe and H{sub 2} in heliummore » recycle gas are less than 1 ppb.« less

Gas-phase detection of solid-state fission product complexes for post-detonation nuclear forensic analysis

DOE PAGES

Stratz, S. Adam; Jones, Steven A.; Oldham, Colton J.; ...

2016-06-27

This study presents the first known detection of fission products commonly found in post-detonation nuclear debris samples using solid sample introduction and a uniquely coupled gas chromatography inductively-coupled plasma time-of-flight mass spectrometer. Rare earth oxides were chemically altered to incorporate a ligand that enhances the volatility of the samples. These samples were injected (as solids) into the aforementioned instrument and detected for the first time. Repeatable results indicate the validity of the methodology, and this capability, when refined, will prove to be a valuable asset for rapid post-detonation nuclear forensic analysis.

Identification of failed fuel element

DOEpatents

Fryer, Richard M.; Matlock, Robert G.

1976-06-22

A passive fission product gas trap is provided in the upper portion of each fuel subassembly in a nuclear reactor. The gas trap consists of an inverted funnel of less diameter than the subassembly having a valve at the apex thereof. An actuating rod extends upwardly from the valve through the subassembly to a point where it can be contacted by the fuel handling mechanism for the reactor. Interrogation of the subassembly for the presence of fission products is accomplished by lowering the fuel handling machine onto the subassembly to press down on the actuating rod and open the valve.

Characterization of fission gas bubbles in irradiated U-10Mo fuel

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

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.

2017-09-01

Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

Gas-phase detection of solid-state fission product complexes for post-detonation nuclear forensic analysis

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

Stratz, S. Adam; Jones, Steven A.; Oldham, Colton J.

This study presents the first known detection of fission products commonly found in post-detonation nuclear debris samples using solid sample introduction and a uniquely coupled gas chromatography inductively-coupled plasma time-of-flight mass spectrometer. Rare earth oxides were chemically altered to incorporate a ligand that enhances the volatility of the samples. These samples were injected (as solids) into the aforementioned instrument and detected for the first time. Repeatable results indicate the validity of the methodology, and this capability, when refined, will prove to be a valuable asset for rapid post-detonation nuclear forensic analysis.

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.

PLENUM INTERIOR, SHOWING ARRANGEMENT OF DRAFT REGULATORS AND FLUES. ...

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

PLENUM INTERIOR, SHOWING ARRANGEMENT OF DRAFT REGULATORS AND FLUES. - Hot Springs National Park, Bathhouse Row, Superior Bathhouse: Mechanical & Piping Systems, State Highway 7, 1 mile north of U.S. Highway 70, Hot Springs, Garland County, AR

1. PLENUM INTERIOR, SHOWING HEATING COILS AND BYPASS Hot ...

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

1. PLENUM INTERIOR, SHOWING HEATING COILS AND BY-PASS - Hot Springs National Park, Bathhouse Row, Fordyce Bathhouse: Mechanical & Piping Systems, State Highway 7, 1 mile north of U.S. Highway 70, Hot Springs, Garland County, AR

Monolithic fuel injector and related manufacturing method

DOEpatents

Ziminsky, Willy Steve [Greenville, SC; Johnson, Thomas Edward [Greenville, SC; Lacy, Benjamin [Greenville, SC; York, William David [Greenville, SC; Stevenson, Christian Xavier [Greenville, SC

2012-05-22

A monolithic fuel injection head for a fuel nozzle includes a substantially hollow vesicle body formed with an upstream end face, a downstream end face and a peripheral wall extending therebetween, an internal baffle plate extending radially outwardly from a downstream end of the bore, terminating short of the peripheral wall, thereby defining upstream and downstream fuel plenums in the vesicle body, in fluid communication by way of a radial gap between the baffle plate and the peripheral wall. A plurality of integral pre-mix tubes extend axially through the upstream and downstream fuel plenums in the vesicle body and through the baffle plate, with at least one fuel injection hole extending between each of the pre-mix tubes and the upstream fuel plenum, thereby enabling fuel in the upstream plenum to be injected into the plurality of pre-mix tubes. The fuel injection head is formed by direct metal laser sintering.

3. PLENUM INTERIOR, SHOWING DRAFT REGULATOR ARRANGEMENT AND STARTER MOTOR. ...

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

3. PLENUM INTERIOR, SHOWING DRAFT REGULATOR ARRANGEMENT AND STARTER MOTOR. - Hot Springs National Park, Bathhouse Row, Lamar Bathhouse: Mechanical & Piping Systems, State Highway 7, 1 mile north of U.S. Highway 70, Hot Springs, Garland County, AR

Photon Intermediate Direct Energy Conversion Using a Strontium-90 Beta Source

NASA Astrophysics Data System (ADS)

Schott, Robert J.

This thesis covers an examination of a need for a compact, long lived power source and a proof of concept for one such design. To begin, tests were done dealing with photovoltaics and their lifetime while undergoing radiation damage from the source of interest, Strontium-90 (Sr-90). After completing these tests a system was designed, built, and ultimately tested over a range of pressures in order to test if a Photon Intermediate Direct Energy Conversion (PIDEC) system would be potentially viable. In brief, the PIDEC system tested for this thesis used two excimer gasses, Argon and Xenon, to produce photons. These gasses were excited into excimer production using a 10 mCi Sr-90 source and held in place at pressures ranging from 10-6 to 2400 psi by a pressure vessel. Photons produced were guided towards a photovoltaic by a mirror chamber lined with high efficiency aluminum mirrors. Outside of the pressure vessel a picoammeter read the current off of the photovoltaic and sent the current to a computer for data processing. Of primary interest was how the current changed based on the amount of energy captured by the gas plenum which was related to the pressure of the system. The overall efficiency of this system was low due to a non-optimized waveguide, much of the beta energy being lost beyond the gas plenum, and other factors. However, the results were sufficient to show that the process was successfully completed and making a new system to optimize for these features is warranted.

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

NASA Astrophysics Data System (ADS)

Barani, T.; Bruschi, E.; Pizzocri, D.; Pastore, G.; Van Uffelen, P.; Williamson, R. L.; Luzzi, L.

2017-04-01

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.

Analysis and comparison of focused ion beam milling and vibratory polishing sample surface preparation methods for porosity study of U-Mo plate fuel for research and test reactors.

PubMed

Westman, Bjorn; Miller, Brandon; Jue, Jan-Fong; Aitkaliyeva, Assel; Keiser, Dennis; Madden, James; Tucker, Julie D

2018-07-01

Uranium-Molybdenum (U-Mo) low enriched uranium (LEU) fuels are a promising candidate for the replacement of high enriched uranium (HEU) fuels currently in use in a high power research and test reactors around the world. Contemporary U-Mo fuel sample preparation uses focused ion beam (FIB) methods for analysis of fission gas porosity. However, FIB possess several drawbacks, including reduced area of analysis, curtaining effects, and increased FIB operation time and cost. Vibratory polishing is a well understood method for preparing large sample surfaces with very high surface quality. In this research, fission gas porosity image analysis results are compared between samples prepared using vibratory polishing and FIB milling to assess the effectiveness of vibratory polishing for irradiated fuel sample preparation. Scanning electron microscopy (SEM) imaging was performed on sections of irradiated U-Mo fuel plates and the micrographs were analyzed using a fission gas pore identification and measurement script written in MatLab. Results showed that the vibratory polishing method is preferentially removing material around the edges of the pores, causing the pores to become larger and more rounded, leading to overestimation of the fission gas porosity size. Whereas, FIB preparation tends to underestimate due to poor micrograph quality and surface damage leading to inaccurate segmentations. Despite the aforementioned drawbacks, vibratory polishing remains a valid method for porosity analysis sample preparation, however, improvements should be made to reduce the preferential removal of material surrounding pores in order to minimize the error in the porosity measurements. Copyright © 2018 Elsevier Ltd. All rights reserved.

Methods and apparatuses for the development of microstructured nuclear fuels

DOEpatents

Jarvinen, Gordon D [Los Alamos, NM; Carroll, David W [Los Alamos, NM; Devlin, David J [Santa Fe, NM

2009-04-21

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

A theory for predicting boundary impedance and resonance frequencies of slotted-wall wind tunnels, including plenum effects

NASA Technical Reports Server (NTRS)

Barger, R. L.

1981-01-01

Wave-induced resonance associated with the geometry of wind-tunnel test sections can occur. A theory that uses acoustic impedance concepts to predict resonance modes in a two dimensional, slotted wall wind tunnel with a plenum chamber is described. The equation derived is consistent with known results for limiting conditions. The computed resonance modes compare well with appropriate experimental data. When the theory is applied to perforated wall test sections, it predicts the experimentally observed closely spaced modes that occur when the wavelength is not long compared with he plenum depth.

A comprehensive cold gas performance study of the Pocket Rocket radiofrequency electrothermal microthruster

NASA Astrophysics Data System (ADS)

Ho, Teck Seng; Charles, Christine; Boswell, Roderick W.

2016-12-01

This paper presents computational fluid dynamics simulations of the cold gas operation of Pocket Rocket and Mini Pocket Rocket radiofrequency electrothermal microthrusters, replicating experiments performed in both sub-Torr and vacuum environments. This work takes advantage of flow velocity choking to circumvent the invalidity of modelling vacuum regions within a CFD simulation, while still preserving the accuracy of the desired results in the internal regions of the microthrusters. Simulated results of the plenum stagnation pressure is in precise agreement with experimental measurements when slip boundary conditions with the correct tangential momentum accommodation coefficients for each gas are used. Thrust and specific impulse is calculated by integrating the flow profiles at the exit of the microthrusters, and are in good agreement with experimental pendulum thrust balance measurements and theoretical expectations. For low thrust conditions where experimental instruments are not sufficiently sensitive, these cold gas simulations provide additional data points against which experimental results can be verified and extrapolated. The cold gas simulations presented in this paper will be used as a benchmark to compare with future plasma simulations of the Pocket Rocket microthruster.

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

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

Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approachmore » to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)« less

Adsorption properties of fission gases Xe and Kr on pristine and doped graphene: A first principle DFT study

NASA Astrophysics Data System (ADS)

Vazhappilly, Tijo; Ghanty, Tapan K.; Jagatap, B. N.

2017-07-01

Graphene has excellent adsorption properties due to large surface area and has been used in applications related to gas sorption and separation. The separation of radioactive noble gases using graphene is an interesting area of research relevant to nuclear waste management. Radioactive noble gases Xe and Kr are present in the off-gas streams from nuclear fission reactors and spent nuclear fuel reprocessing plants. The entrapment of these volatile fission gases is important in the context of nuclear safety. The separation of Xe from Kr is extremely difficult, and energy intensive cryogenic distillation is generally employed. Physisorption based separation techniques using porous materials is a cost effective alternative to expensive cryogenic distillation. Thus, adsorption of noble gases on graphene is relevant for fundamental understanding of physisorption process. The properties of graphene can be tuned by doping and incorporation of defects. In this regard, we study the binding affinity of Xe and Kr in pristine and doped graphene sheets. We employ first principle calculations using density functional theory, corrected for dispersion interactions. The structural parameters obtained from the current study show excellent agreement with the available theoretical and experimental observations on similar systems. Noble gas adsorption energies on pristine graphene match very well with the available literature. Our results show that the binding energy of fission gases Xe and Kr on graphene can be considerably improved through doping the lattice with a heteroatom.

Linking photochemistry in the gas and solution phase: S-H bond fission in p-methylthiophenol following UV photoexcitation.

PubMed

Oliver, Thomas A A; Zhang, Yuyuan; Ashfold, Michael N R; Bradforth, Stephen E

2011-01-01

Gas-phase H (Rydberg) atom photofragment translational spectroscopy and solution-phase femtosecond-pump dispersed-probe transient absorption techniques are applied to explore the excited state dynamics of p-methylthiophenol connecting the short time reactive dynamics in the two phases. The molecule is excited at a range of UV wavelengths from 286 to 193 nm. The experiments clearly demonstrate that photoexcitation results in S-H bond fission--both in the gas phase and in ethanol solution-and that the resulting p-methythiophenoxyl radical fragments are formed with significant vibrational excitation. In the gas phase, the recoil anisotropy of the H atom and the vibrational energy disposal in the p-MePhS radical products formed at the longer excitation wavelengths reveal the operation of two excited state dissociation mechanisms. The prompt excited state dissociation motif appears to map into the condensed phase also. In both phases, radicals are produced in both their ground and first excited electronic states; characteristic signatures for both sets of radical products are already apparent in the condensed phase studies after 50 fs. No evidence is seen for either solute ionisation or proton coupled electron transfer--two alternate mechanisms that have been proposed for similar heteroaromatics in solution. Therefore, at least for prompt S-H bond fissions, the direct observation of the dissociation process in solution confirms that the gas phase photofragmentation studies indeed provide important insights into the early time dynamics that transfer to the condensed phase.

System of treating flue gas

DOEpatents

Ziegler, D.L.

1975-12-01

A system is described for treating or cleaning incinerator flue gas containing acid gases and radioactive and fissionable contaminants. Flue gas and a quench solution are fed into a venturi and then tangentially into the lower portion of a receptacle for restricting volumetric content of the solution. The upper portion of the receptacle contains a scrub bed to further treat or clean the flue gas.

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.

Thermal-stress fracture and fractography in UO/sub 2/

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

Kennedy, C.R.; Bandyopadhyay, G.

1976-01-01

Pressed and sintered UO/sub 2/ pellets were thermally shocked by quenching into a water bath at room temperature. The cracking behavior and strength degradation, as measured by the diametral compression technique, in these quench tests are discussed. Fractography of the thermally shocked specimens by scanning-electron microscopy indicated predominantly intergranular fracture in UO/sub 2/ in severe thermal-shock tests. The implication of this observation is that intergranular cracking may occur during the initial heat up in a reactor. Because fission gas bubbles tend to migrate toward the grain boundary, preferential microcracking along the boundary may strongly affect subsequent fission gas release behavior.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

NASA Astrophysics Data System (ADS)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Fission fragment assisted reactor concept for space propulsion: Foil reactor

NASA Technical Reports Server (NTRS)

Wright, Steven A.

1991-01-01

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

30 CFR 581.8 - Rights to minerals.

Code of Federal Regulations, 2013 CFR

2013-07-01

... THAN OIL, GAS, AND SULPHUR IN THE OUTER CONTINENTAL SHELF General § 581.8 Rights to minerals. (a... leases; (2) Oil; (3) Gas; (4) Sulphur; (5) Minerals produced in direct association with oil, gas, or... essential to production of fissionable materials which are reserved pursuant to section 12(a) of the Act. (b...

30 CFR 581.8 - Rights to minerals.

Code of Federal Regulations, 2012 CFR

2012-07-01

... THAN OIL, GAS, AND SULPHUR IN THE OUTER CONTINENTAL SHELF General § 581.8 Rights to minerals. (a... leases; (2) Oil; (3) Gas; (4) Sulphur; (5) Minerals produced in direct association with oil, gas, or... essential to production of fissionable materials which are reserved pursuant to section 12(a) of the Act. (b...

30 CFR 581.8 - Rights to minerals.

Code of Federal Regulations, 2014 CFR

2014-07-01

... THAN OIL, GAS, AND SULPHUR IN THE OUTER CONTINENTAL SHELF General § 581.8 Rights to minerals. (a... leases; (2) Oil; (3) Gas; (4) Sulphur; (5) Minerals produced in direct association with oil, gas, or... essential to production of fissionable materials which are reserved pursuant to section 12(a) of the Act. (b...

Innovative nuclear thermal propulsion technology evaluation - Results of the NASA/DOE task team study

NASA Technical Reports Server (NTRS)

Howe, Steven D.; Borowski, Stanley; Motloch, Chet; Helms, Ira; Diaz, Nils; Anghaie, Samim; Latham, Thomas

1991-01-01

In response to findings from two NASA/DOE nuclear propulsion workshops, six task teams were created to continue evaluation of various propulsion concepts, from which evolved an innovative concepts subpanel to evaluate thermal propulsion concepts which did not utilize solid fuel. This subpanel endeavored to evaluate each concept on a level technology basis, and to identify critical issues, technologies, and early proof-of-concept experiments. Results of the concept studies including the liquid core fission, the gas core fission, the fission foil reactors, explosively driven systems, fusion, and antimatter are presented.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Investigations of the Application of CFD to Flow Expected in the Lower Plenum of the Prismatic VHTR

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

Richard W.Johnson; Tara Gallaway; Donna P. Guillen

2006-09-01

The Generation IV (Gen IV) very high temperature reactor (VHTR) will either be a prismatic (block) or pebble bed design. However, a prismatic VHTR reference design, based on the General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) [General Atomics, 1996] has been developed for preliminary analysis purposes [MacDonald, et al., 2003]. Numerical simulation studies reported herein are based on this reference design. In the lower plenum of the prismatic reference design, the flow will be introduced by dozens of turbulent jets from the core above. The jet flow will encounter rows of columns that support the core. The flow from themore » core will have to turn ninety degrees and flow toward the exit duct as it passed through the forest of support columns. Due to the radial variation of the power density in the core, the jets will be at various temperatures at the inlet to the lower plenum. This presents some concerns, including that local hot spots may occur in the lower plenum. This may have a deleterious effect on the materials present as well as cause a variation in temperature to be present as the flow enters the power conversion system machinery, which could cause problems with the operation of the machinery. In the past, systems analysis codes have been used to model flow in nuclear reactor systems. It is recognized, however, that such codes are not capable of modeling the local physics of the flow to be able to analyze for local mixing and temperature variations. This has led to the determination that computational fluid dynamic (CFD) codes be used, which are generally regarded as having the capability of accurately simulating local flow physics. Accurate flow modeling involves determining appropriate modeling strategies needed to obtain accurate analyses. These include determining the fineness of the grid needed, the required iterative convergence tolerance, which numerical discretization method to use, and which turbulence model and wall treatment should be employed. It also involves validating the computer code and turbulence model against a series of separate and combined flow phenomena and selecting the data used for the validation. This report describes progress made to identify proper modeling strategies for simulating the lower plenum flow for the task entitled “CFD software validation of jets in crossflow,” which was designed to investigate the issues pertaining to the validation process. The flow phenomenon previously chosen to investigate is flow in a staggered tube bank because it is shown by preliminary simulations to be the location of the highest turbulence intensity in the lower plenum Numerical simulations were previously obtained assuming that the flow is steady. Various turbulence models were employed along with strategies to reduce numerical error to allow appropriate comparisons of the results. It was determined that the sophisticated Reynolds stress model (RSM) provided the best results. It was later determined that the flow is an unsteady flow wherein circulating eddies grow behind the tube and ‘peel off’ alternately from the top and the bottom of the tube. Additional calculations show that the mean velocity is well predicted when the flow is modeled as an unsteady flow. The results for U are clearly superior for the unsteady computations; the unsteady computations for the turbulence stress are similar to those for the steady calculations, showing the same trends. It is clear that strategie« less

Deformation and the timing of gas generation and migration in the eastern Brooks Range foothills, Arctic National Wildlife Refuge, Alaska

USGS Publications Warehouse

Parris, T.M.; Burruss, R.C.; O'Sullivan, P. B.

2003-01-01

Along the southeast border of the 1002 Assessment Area in the Arctic National Wildlife Refuge, Alaska, an explicit link between gas generation and deformation in the Brooks Range fold and thrust belt is provided through petrographic, fluid inclusion, and stable isotope analyses of fracture cements integrated with zircon fission-track data. Predominantly quartz-cemented fractures, collected from thrusted Triassic and Jurassic rocks, contain crack-seal textures, healed microcracks, and curved crystals and fluid inclusion populations, which suggest that cement growth occurred before, during, and after deformation. Fluid inclusion homogenization temperatures (175-250??C) and temperature trends in fracture samples suggest that cements grew at 7-10 km depth during the transition from burial to uplift and during early uplift. CH4-rich (dry gas) inclusions in the Shublik Formation and Kingak Shale are consistent with inclusion entrapment at high thermal maturity for these source rocks. Pressure modeling of these CH4-rich inclusions suggests that pore fluids were overpressured during fracture cementation. Zircon fission-track data in the area record postdeposition denudation associated with early Brooks Range deformation at 64 ?? 3 Ma. With a closure temperature of 225-240??C, the zircon fission-track data overlap homogenization temperatures of coeval aqueous inclusions and inclusions containing dry gas in Kingak and Shublik fracture cements. This critical time-temperature relationship suggests that fracture cementation occurred during early Brooks Range deformation. Dry gas inclusions suggest that Shublik and Kingak source rocks had exceeded peak oil and gas generation temperatures at the time structural traps formed during early Brooks Range deformation. The timing of hydrocarbon generation with respect to deformation therefore represents an important exploration risk for gas exploration in this part of the Brooks Range fold and thrust belt. The persistence of gas high at thermal maturity levels suggests, however, that significant volumes of gas may have been generated.

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

2002-08-01

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

NASA Technical Reports Server (NTRS)

Thoms, K. R.

1975-01-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

Model investigation of inlet plenum flow straightening techniques for altitude test facility

NASA Technical Reports Server (NTRS)

Riddlebaugh, S. M.; Linke, H. G.

1976-01-01

An investigation was conducted to evaluate and improve the quality of the airflow to be supplied to the engine in altitude test chambers 3 and 4 of the Propulsion Systems Laboratory at the Lewis Research Center. One-twentieth-scale models of the inlet plenum chamber of the two test chambers were used in the investigation to minimize time and cost. It was possible to reduce the velocity spread in the inlet plenum from approximately 100 m/sec (330 ft/sec) to approximately 10 m/sec (30 ft/sec) through the combined use of flow diverters, multiple spaced screens, flow straighteners, and turning vanes.

Tuned intake air inlet for a rotary engine

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

Corbett, W.D.; Sheaffer, B.L.

This patent describes, in a rotary internal combustion engine, an improved assembly for providing a balanced flow of combustion air to the fuel supply inlet. It comprises: a plenum chamber attached to the engine block, the plenum chamber including an air inlet adapted to receive air from the cooling air exit passage and an air outlet for the discharge of air; and an outlet conduit connecting the air outlet and the fuel supply inlet. The conduit disposed to partially surround the plenum chamber to provide a conduit length substantially greater than the distance from the cooling air exit passage totmore » he fuel supply inlet.« less

Development of a novel passive top-down uniflow scavenged two-stroke GDI engine

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

Ciccarelli, G.; Reynolds, Steve; Oliver, Phillip

2010-02-15

The design and performance characteristics of a novel top-down uniflow scavenged gasoline direct-injection two-stroke engine are presented. The novelty of the engine lies in the cylinder head that contains multiple check valves that control scavenging airflow into the cylinder from a supercharged air plenum. When the cylinder pressure drops below the intake plenum pressure during the expansion stroke, air flows into the cylinder through the check valves. During compression the cylinder pressure increases to a level above the intake plenum pressure and the check valves close preventing back-flow into the intake plenum. The engine head design provides asymmetrical intake valvemore » timing without the use of poppet valves and the associated valve-train. In combination with an external Roots-type supercharger that supplies the plenum and exhaust ports at the bottom of the cylinder wall, the novel head provides top-down uniflow air scavenging. Motoring tests indicated that the check valves seal and the peak pressure is governed by the compression ratio. The only drawback observed is that valve closing is delayed as the engine speed increases. In order to investigate the valve dynamics, additional tests were performed in an optically-accessible cold flow test rig that enabled the direct measurement of valve opening and closing time under various conditions. (author)« less

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

NASA Technical Reports Server (NTRS)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

Effect of point defects on the thermal conductivity of UO2: molecular dynamics simulations

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

Liu, Xiang-Yang; Stanek, Christopher Richard; Andersson, Anders David Ragnar

2015-07-21

The thermal conductivity of uranium dioxide (UO 2) fuel is an important materials property that affects fuel performance since it is a key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. [1] The thermal conductivity of UO 2 nuclear fuel is also affected by fission gas, fission products, defects, and microstructural features such as grain boundaries. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of irradiation induced point defects on the thermal conductivity of UO 2, as a function of defectmore » concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel [2].« less

Application of NASTRAN to a fluid solids unit in the petroleum industry. [plenum/cyclone/dipleg assembly

NASA Technical Reports Server (NTRS)

Nelson, N. W.

1975-01-01

The application of NASTRAN to the design of a fluid solids unit plenum/cyclone/dipleg assembly is described. The major loads considered are thermal, pressure, and gravity. Such applications are of interest in the petroleum industry since the equipment described is historically critical.

Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

2016-06-01

Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

Apparatus and methods for relieving thermally induced stresses in inner and outer bands of thermally cooled turbine nozzle stages

DOEpatents

Yu, Yufeng Phillip; Itzel, Gary Michael; Correia, Victor H. S.

2002-01-01

To control the temperature mismatch between the inner and outer bands and covers forming plenums with the inner and outer bands on sides thereof remote from the hot gas path, passages extend from the leading edge of the covers in communication with the hot gases of combustion to the trailing edge of the covers in communication with the hot gas flowpath. A mixing chamber is provided in each passage in communication with compressor discharge air for mixing the hot gases of combustion and compressor discharge air for flow through the passage, thereby heating the cover and minimizing the temperature differential between the inner and outer bands and their respective covers. The passages are particularly useful adjacent the welded or brazed joints between the covers and inner band portions.

Enrichment, separation, and gas-chromatographic and mass-spectrometric analyses of fission products from irradiated or heated fats, oils, and test substances (in German)

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

Beck, B.

1973-01-01

From international colloquium: the identification of irradiated foodstuffs; Karlsrahe, Germany (24 Oct 1973). Tripalmitate, tristearate, trioleate, oleic acid methyl ester, linoleic acid methyl ester, lauric acid, lard, coconut butter, sunflower oil, and olive oil were irradiated at 0.5-6 Mrad,or heated up to 174 deg C for 24 hr. The fission products were fractionally distilled with silica gel according to polarity into elutropic series. Subsequent identification and quantitative determination were done by gas chromatography and mass spectrometry. Approximately 28 hydrocarbons and 24 oxygen compounds are dealt with, the typical substances being described individually as regards their identification and quantitative distribution. (GE)

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, D.M.; Nesbitt, L.B.

1997-01-21

A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

Adsorption and excess fission Xe - Adsorption of Xe on vacuum crushed minerals

NASA Technical Reports Server (NTRS)

Bernatowicz, T. J.; Kramer, F. E.; Podosek, F. A.; Honda, M.

1982-01-01

It is hypothesized that adsorption is not likely to provide a sufficiently precise mechanism for the concentration of excess fission Xe in the entire lunar regolith, in view of laboratory analogs of the lunar soil and calculations of the residence times of noble gases in the present day regolith. Lunar cold trap and episodic degassing models are difficult to reconcile, however, with the generality of excess fission Xe in all gas-rich highland breccias. It is concluded that the high Xe concentration in such highland breccias is not the result of Xe adsorption prior to the trapping of this component.

A MODEL FOR FISSION-GAS RELEASE FROM POROUS FUELS IN LOW-PERMEABILITY CONTAINERS

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

Prados, J.W.

1961-08-25

A simple mathematical model was developed to describe the steady-state release rate of gaseous fission products from porous ceramic fuels in low- permeability containers. The resulting equations are used to analyze experimental release rate results obtained from a UC/sub 2/-fueled graphite fuel body enclosed in a low-permeability impregnated graphite container. The relative release rates of the fission-product species Kr/sup 85m/, Kr/sup 88/, and Xe/sup 133/ were predicted with reasonable success. Absolute-rate predictions were not possible due to lack of information on true permeability and porosity profiles in the graphite container. (auth)

Isolation and Purification of the Xenon Fraction of 252Cf Spontaneous Fission Products for the Production of Radio Xenon Calibration Standards

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

McGrath, Christopher A.

2015-04-01

The presence of radioactive xenon isotopes indicates that fission events have occurred, and is used to help enforce the Comprehensive Test Ban Treaty. Idaho National Laboratory (INL) produces 135Xe, 133mXe, 133Xe, and 131mXe standards used for the calibration and testing of collection equipment and analytical techniques used to monitor radio xenon emissions. At INL, xenon is produced and collected as one of several spontaneous fission products from a 252Cf source. Further chromatographic purification of the fission gases ensures the separations of the xenon fraction for selective collection. An explanation of the fission gas collection, separation and purification is presented. Additionally,more » the range of 135Xe to 133Xe ratio that can be isolated is explained. This is an operational update on the work introduced previously, now that it is in operation and has been recharged with a second 252Cf source.« less

Cascading for Large Effective Third Order Nonlinearities

DTIC Science & Technology

1997-12-18

edited by T. Tamir, H. Bertoni and G. Griffel , (Plenum Press, New York, 1995), p381-9 9. G.I. Stegeman, R. Schiek, G. Krijnen, W. Torruellas, M...Guided Wave Optoelectronics, edited by T. Tamir, H. Bertoni and G. Griffel , (Plenum Press, New York, 1995), 371-9. 10. W.E. Torruellas, Z. Wang, L

Gaseous swelling of U 3 Si 2 during steady-state LWR operation: A rate theory investigation

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

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Rate theory simulations of fission gas behavior in U 3Si 2 are reported for light water reactor (LWR) steady-state operation scenarios. We developed a model of U 3Si 2 and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U 3Si 2 swelling behavior between 390 K and 1190 K. Under typical steady-state LWR operating conditions where U 3Si 2 temperature is expected to be below 1000 K,more » intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U 3Si 2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.« less

Gaseous swelling of U 3 Si 2 during steady-state LWR operation: A rate theory investigation

DOE PAGES

Miao, Yinbin; Gamble, Kyle A.; Andersson, David; ...

2017-07-25

Rate theory simulations of fission gas behavior in U 3Si 2 are reported for light water reactor (LWR) steady-state operation scenarios. We developed a model of U 3Si 2 and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U 3Si 2 swelling behavior between 390 K and 1190 K. Under typical steady-state LWR operating conditions where U 3Si 2 temperature is expected to be below 1000 K,more » intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U 3Si 2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.« less

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

Liang, Linyun; Mei, Zhi-Gang; Yacout, Abdellatif M.

We have developed a mesoscale phase-field model for studying the effect of recrystallization on the gas-bubble-driven swelling in irradiated U-Mo alloy fuel. The model can simulate the microstructural evolution of the intergranular gas bubbles on the grain boundaries as well as the recrystallization process. Our simulation results show that the intergranular gas-bubble-induced fuel swelling exhibits two stages: slow swelling kinetics before recrystallization and rapid swelling kinetics with recrystallization. We observe that the recrystallization can significantly expedite the formation and growth of gas bubbles at high fission densities. The reason is that the recrystallization process increases the nucleation probability of gasmore » bubbles and reduces the diffusion time of fission gases from grain interior to grain boundaries by increasing the grain boundary area and decreasing the diffusion distance. The simulated gas bubble shape, size distribution, and density on the grain boundaries are consistent with experimental measurements. We investigate the effect of the recrystallization on the gas-bubble-driven fuel swelling in UMo through varying the initial grain size and grain aspect ratio. We conclude that the initial microstructure of fuel, such as grain size and grain aspect ratio, can be used to effectively control the recrystallization and therefore reduce the swelling in U-Mo fuel.« less

Building a multi-cathode-gas-filled scintillator detector for fission fragments

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

Mahgoub, M., E-mail: mmahgoub@jazanu.edu.sa; Physics department, Technical University of Munich, D-85748 Garching

2016-06-10

Radiation cannot be detected directly by human senses, indeed detecting and identifying the fission products or decay yield with high accuracy is a great challenge for experimental physicist. In this work we are building a Multi-Cathode-Gas-filled Scintillator MCGS detector. The detector consists of two parts. First: anode-wire proportional chamber and cathode strip foil, which measure the energy loss of the particles in the gas, due to the ionization, and identifies the position of the products on the detector plane depending on their energy with the presence of a magnetic field. Second: a 7 mm thick scintillator attached to a photomultipliermore » tube in the back end of the detector. This part measures the rest energy of the particles. A data acquisition system records the events and the particles infonnation. The yields are identified from the energy loss to rest energy ratio.« less

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

NASA Technical Reports Server (NTRS)

Kazeminezhad, F.; Anghaie, S.

2008-01-01

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

DOE PAGES

Barani, T.; Bruschi, E.; Pizzocri, D.; ...

2017-01-03

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which ismore » applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.« less

Diffusion of Zr, Ru, Ce, Y, La, Sr and Ba fission products in UO 2

DOE PAGES

Perriot, R.; Liu, X. -Y.; Stanek, C. R.; ...

2015-01-08

The diffusivity of the solid fission products (FP) Zr (Zr 4+), Ru (Ru 4+, Ru 3+), Ce (Ce 4+), Y (Y 3+), La (La 3+), Sr (Sr 2+) and Ba (Ba 2+) by a vacancy mechanism has been calculated, using a combination of density functional theory (DFT) and empirical potential (EP) calculations. The activation energies for the solid fission products are compared to the activation energy for Xe fission gas atoms calculated previously. Apart from Ru, the solid fission products all exhibit higher activation energy than Xe. Furthermore, for all solid FPs except Y 3+, the migration of the FPmore » has lower barrier than the migration of a neighboring U atom, making the latter the rate limiting step for direct migration. An indirect mechanism, consisting of two successive migrations around the FP, is also investigated. The calculated diffusivities show that most solid fission products diffuse with rates similar to U self-diffusion. But, Ru, Ba and Sr exhibit faster diffusion than the other solid FPs, with Ru 3+ and Ru 4+ diffusing even faster than Xe for T < 1200 K. The diffusivities correlate with the observed fission product solubility in UO 2, and the tendency to form metallic and oxide second phase inclusions.« less

HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

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

Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.

High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Testmore » Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two primary goals. First, the test was intended to assess the retention of fission products in loose kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material comprising the compact. Second, this test served as an evaluation of the FACS fission product condensation plate collection efficiency.« less

Enhanced heat transfer surface for cast-in-bump-covered cooling surfaces and methods of enhancing heat transfer

DOEpatents

Chiu, Rong-Shi Paul; Hasz, Wayne Charles; Johnson, Robert Alan; Lee, Ching-Pang; Abuaf, Nesim

2002-01-01

An annular turbine shroud separates a hot gas path from a cooling plenum containing a cooling medium. Bumps are cast in the surface on the cooling side of the shroud. A surface coating overlies the cooling side surface of the shroud, including the bumps, and contains cooling enhancement material. The surface area ratio of the cooling side of the shroud with the bumps and coating is in excess of a surface area ratio of the cooling side surface with bumps without the coating to afford increased heat transfer across the element relative to the heat transfer across the element without the coating.

NEUTRONIC REACTOR FUEL PUMP

DOEpatents

Cobb, W.G.

1959-06-01

A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

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

Pastore, Giovanni; Rabiti, Cristian; Pizzocri, Davide

PolyPole is a numerical algorithm for the calculation of intra-granular fission gas release. In particular, the algorithm solves the gas diffusion problem in a fuel grain in time-varying conditions. The program has been extensively tested. PolyPole combines a high accuracy with a high computational efficiency and is ideally suited for application in fuel performance codes.

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.

Multi-purpose wind tunnel reaction control model block

NASA Technical Reports Server (NTRS)

Dresser, H. S.; Daileda, J. J. (Inventor)

1978-01-01

A reaction control system nozzle block is provided for testing the response characteristics of space vehicles to a variety of reaction control thruster configurations. A pressurized air system is connected with the supply lines which lead to the individual jet nozzles. Each supply line terminates in a compact cylindrical plenum volume, axially perpendicular and adjacent to the throat of the jet nozzle. The volume of the cylindrical plenum is sized to provide uniform thrust characteristics from each jet nozzle irrespective of the angle of approach of the supply line to the plenum. Each supply line may be plugged or capped to stop the air supply to selected jet nozzles, thereby enabling a variety of nozzle configurations to be obtained from a single model nozzle block.

Experiments in a three-dimensional adaptive-wall wind tunnel

NASA Technical Reports Server (NTRS)

Schairer, E. T.

1983-01-01

Three dimensional adaptive-wall experiments were performed in the Ames Research Center (ARC) 25- by 13-cm indraft wind tunnel. A semispan wing model was mounted to one sidewall of a test section with solid sidewalls, and slotted top and bottom walls. The test section had separate top and bottom plenums which were divided into streamwise and cross-stream compartments. An iterative procedure was demonstrated for measuring wall interference and for adjusting the plenum compartment pressures to eliminate such interference. The experiments were conducted at a freestream Mach number of 0.60 and model angles of attack between 0 and 6 deg. Although in all the experiments wall interference was reduced after the plenum pressures were adjusted, interference could not be completely eliminated.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.

1973-01-01

The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

Tables of critical-flow functions and thermodynamic properties for methane and computational procedures for both methane and natural gas

NASA Technical Reports Server (NTRS)

Johnson, R. C.

1972-01-01

Procedures for calculating the mass flow rate of methane and natural gas through nozzles are given, along with the FORTRAN 4 subroutines used to make these calculations. Three sets of independent variables are permitted in these routines. In addition to the plenum pressure and temperature, the third independent variable is either nozzle exit pressure, Mach number, or temperature. A critical-flow factor that becomes a convenient means for determining the mass flow rate of methane through critical-flow nozzles is tabulated. Other tables are included for nozzle throat velocity and critical pressure, density, and temperature ratios, along with some thermodynamic properties of methane, including compressibility factor, enthalpy, entropy, specific heat, specific-heat ratio, and speed of sound. These tabulations cover a temperature range from 120 to 600 K and pressures to 3 million N/sq m.

The DART dispersion analysis research tool: A mechanistic model for predicting fission-product-induced swelling of aluminum dispersion fuels. User`s guide for mainframe, workstation, and personal computer applications

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

Rest, J.

1995-08-01

This report describes the primary physical models that form the basis of the DART mechanistic computer model for calculating fission-product-induced swelling of aluminum dispersion fuels; the calculated results are compared with test data. In addition, DART calculates irradiation-induced changes in the thermal conductivity of the dispersion fuel, as well as fuel restructuring due to aluminum fuel reaction, amorphization, and recrystallization. Input instructions for execution on mainframe, workstation, and personal computers are provided, as is a description of DART output. The theory of fission gas behavior and its effect on fuel swelling is discussed. The behavior of these fission products inmore » both crystalline and amorphous fuel and in the presence of irradiation-induced recrystallization and crystalline-to-amorphous-phase change phenomena is presented, as are models for these irradiation-induced processes.« less

3D modeling of missing pellet surface defects in BWR fuel

DOE PAGES

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.; ...

2016-07-26

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Experimental evaluation of blockage ratio and plenum evacuation system flow effects on pressure distribution for bodies of revolution in 0.1 scale model test section of NASA Lewis Research Center's proposed altitude wind tunnel

NASA Technical Reports Server (NTRS)

Burley, Richard R.; Harrington, Douglas E.

1987-01-01

An experimental investigation was conducted in the slotted test section of the 0.1-scale model of the proposed Altitude Wind Tunnel to evaluate wall interference effects at tunnel Mach numbers from 0.70 to 0.95 on bodies of revolution with blockage rates of 0.43, 3, 6, and 12 percent. The amount of flow that had to be removed from the plenum chamber (which surrounded the slotted test section) by the plenum evacuation system (PES) to eliminate wall interference effects was determined. The effectiveness of tunnel reentry flaps in removing flow from the plenum chamber was examined. The 0.43-percent blockage model was the only one free of wall interference effects with no PES flow. Surface pressures on the forward part of the other models were greater than interference-free results and were not influenced by PES flow. Interference-free results were achieved on the aft part of the 3- and 6-percent blockage models with the proper amount of PES flow. The required PES flow was substantially reduced by opening the reentry flaps.

VICTORIA-92 pretest analyses of PHEBUS-FPT0

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

Bixler, N.E.; Erickson, C.M.

FPT0 is the first of six tests that are scheduled to be conducted in an experimental reactor in Cadarache, France. The test apparatus consists of an in-pile fuel bundle, an upper plenum, a hot leg, a steam generator, a cold leg, and a small containment. Thus, the test is integral in the sense that it attempts to simulate all of the processes that would be operative in a severe nuclear accident. In FPT0, the fuel will be trace irradiated; in subsequent tests high burn-up fuel will be used. This report discusses separate pretest analyses of the FPT0 fuel bundle andmore » primary circuit have been conducted using the USNRC`s source term code, VICTORIA-92. Predictions for release of fission product, control rod, and structural elements from the test section are compared with those given by CORSOR-M. In general, the releases predicted by VICTORIA-92 occur earlier than those predicted by CORSOR-M. The other notable difference is that U release is predicted to be on a par with that of the control rod elements; CORSOR-M predicts U release to be about 2 orders of magnitude greater.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

AMTEC vapor-vapor series connected cells

NASA Technical Reports Server (NTRS)

Underwood, Mark L. (Inventor); Williams, Roger M. (Inventor); Ryan, Margaret A. (Inventor); Nakamura, Barbara J. (Inventor); Oconnor, Dennis E. (Inventor)

1995-01-01

An alkali metal thermoelectric converter (AMTEC) having a plurality of cells structurally connected in series to form a septum dividing a plenum into two chambers, and electrically connected in series, is provided with porous metal anodes and porous metal cathodes in the cells. The cells may be planar or annular, and in either case a metal alkali vapor at a high temperature is provided to the plenum through one chamber on one side of the wall and returned to a vapor boiler after condensation at a chamber on the other side of the wall in the plenum. If the cells are annular, a heating core may be placed along the axis of the stacked cells. This arrangement of series-connected cells allows efficient generation of power at high voltage and low current.

Micro-fluid exchange coupling apparatus

NASA Technical Reports Server (NTRS)

Johnson, J. E., Jr.; Swartz, P. F. (Inventor)

1980-01-01

In a macro-fluid exchange, a hollow needle, such as a syringe needle, is provided for penetrating the fluid conduit of the animal. The syringe needle is coupled to a plenum chamber having an inlet and outlet port. The plenum chamber is coupled to the syringe needle via the intermediary of a standard quick disconnect coupling fitting. The plenum chamber is carried at the end of a drive rod which is coupled to a micrometer drive head. The micrometer drive head is slidably and pivotably coupled to a pedestal for adjusting the height and angle of inclination of the needle relative to a reference base support. The needle is positioned adjacent to the incised trachea or a blood vessel of a small animal and the micrometer drive head is operated for penetrating the fluid conduit of the animal.

Spontaneous and induced emission of XeCl* excimer molecules under pumping of Xe - CCl4 and Ar - Xe - CCl4 gas mixtures with a low CCl4 content by fast electrons and uranium fission fragments

NASA Astrophysics Data System (ADS)

Mis'kevich, A. I.; Guo, J.; Dyuzhov, Yu A.

2013-11-01

The spontaneous and induced emission of XeCl* excimer molecules upon excitation of Xe - CCl4 and Ar - Xe - CCl4 gas mixtures with a low CCl4 content by high-energy charged particles [a pulsed high-energy electron beam and products of neutron nuclear reaction 235U(n, f)] has been experimentally studied. The electron energy was 150 keV, and the pump current pulse duration and amplitude were 5 ns and 5 A, respectively. The energy of fission fragments did not exceed 100 MeV, the duration of the neutron pump pulse was 200 μs, and the specific power contribution to the gas was about 300 W cm-3. Electron beam pumping in a cell 4 cm long with a cavity having an output mirror transmittance of 2.7% gives rise to lasing on the B → X transition in the XeCl* molecule (λ = 308 nm) with a gain α = 0.0085 cm-1 and fluorescence efficiency η ≈ 10%. Pumping by fission fragments in a 250-cm-long cell with a cavity formed by a highly reflecting mirror and a quartz window implements amplified spontaneous emission (ASE) with an output power of 40 - 50 kW sr-1 and a base ASE pulse duration of ~200 ms.

ZPPR-20 phase D : a cylindrical assembly of polyethylene moderated U metal reflected by beryllium oxide and polyethylene.

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

Lell, R.; Grimm, K.; McKnight, R.

The Zero Power Physics Reactor (ZPPR) fast critical facility was built at the Argonne National Laboratory-West (ANL-W) site in Idaho in 1969 to obtain neutron physics information necessary for the design of fast breeder reactors. The ZPPR-20D Benchmark Assembly was part of a series of cores built in Assembly 20 (References 1 through 3) of the ZPPR facility to provide data for developing a nuclear power source for space applications (SP-100). The assemblies were beryllium oxide reflected and had core fuel compositions containing enriched uranium fuel, niobium and rhenium. ZPPR-20 Phase C (HEU-MET-FAST-075) was built as the reference flight configuration.more » Two other configurations, Phases D and E, simulated accident scenarios. Phase D modeled the water immersion scenario during a launch accident, and Phase E (SUB-HEU-MET-FAST-001) modeled the earth burial scenario during a launch accident. Two configurations were recorded for the simulated water immersion accident scenario (Phase D); the critical configuration, documented here, and the subcritical configuration (SUB-HEU-MET-MIXED-001). Experiments in Assembly 20 Phases 20A through 20F were performed in 1988. The reference water immersion configuration for the ZPPR-20D assembly was obtained as reactor loading 129 on October 7, 1988 with a fissile mass of 167.477 kg and a reactivity of -4.626 {+-} 0.044{cents} (k {approx} 0.9997). The SP-100 core was to be constructed of highly enriched uranium nitride, niobium, rhenium and depleted lithium. The core design called for two enrichment zones with niobium-1% zirconium alloy fuel cladding and core structure. Rhenium was to be used as a fuel pin liner to provide shut down in the event of water immersion and flooding. The core coolant was to be depleted lithium metal ({sup 7}Li). The core was to be surrounded radially with a niobium reactor vessel and bypass which would carry the lithium coolant to the forward inlet plenum. Immediately inside the reactor vessel was a rhenium baffle which would act as a neutron curtain in the event of water immersion. A fission gas plenum and coolant inlet plenum were located axially forward of the core. Some material substitutions had to be made in mocking up the SP-100 design. The ZPPR-20 critical assemblies were fueled by 93% enriched uranium metal because uranium nitride, which was the SP-100 fuel type, was not available. ZPPR Assembly 20D was designed to simulate a water immersion accident. The water was simulated by polyethylene (CH{sub 2}), which contains a similar amount of hydrogen and has a similar density. A very accurate transformation to a simplified model is needed to make any of the ZPPR assemblies a practical criticality-safety benchmark. There is simply too much geometric detail in an exact model of a ZPPR assembly, particularly as complicated an assembly as ZPPR-20D. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation will be described in a later section. First, Assembly 20D was modeled in full detail--every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from this model were converted to an RZ model. ZPPR Assembly 20D has been determined to be an acceptable criticality-safety benchmark experiment.« less

Aerothermodynamic Design Sensitivities for a Reacting Gas Flow Solver on an Unstructured Mesh Using a Discrete Adjoint Formulation

NASA Astrophysics Data System (ADS)

Thompson, Kyle Bonner

An algorithm is described to efficiently compute aerothermodynamic design sensitivities using a decoupled variable set. In a conventional approach to computing design sensitivities for reacting flows, the species continuity equations are fully coupled to the conservation laws for momentum and energy. In this algorithm, the species continuity equations are solved separately from the mixture continuity, momentum, and total energy equations. This decoupling simplifies the implicit system, so that the flow solver can be made significantly more efficient, with very little penalty on overall scheme robustness. Most importantly, the computational cost of the point implicit relaxation is shown to scale linearly with the number of species for the decoupled system, whereas the fully coupled approach scales quadratically. Also, the decoupled method significantly reduces the cost in wall time and memory in comparison to the fully coupled approach. This decoupled approach for computing design sensitivities with the adjoint system is demonstrated for inviscid flow in chemical non-equilibrium around a re-entry vehicle with a retro-firing annular nozzle. The sensitivities of the surface temperature and mass flow rate through the nozzle plenum are computed with respect to plenum conditions and verified against sensitivities computed using a complex-variable finite-difference approach. The decoupled scheme significantly reduces the computational time and memory required to complete the optimization, making this an attractive method for high-fidelity design of hypersonic vehicles.

SEPARATION APPARATUS

DOEpatents

Huff, J.B.

1962-03-13

A furnace apparatus is designed for treating a nuclear reactor waste solution. The solution is sprayed onto a bed of burning petroleum coke which expels water, the more volatile fission products, and nitrogen oxides. Next, chlorine gas is introduced from below which causes aluminum to volatilize as aluminum chloride and along with it certain fission products including Nb/sup 95/ and Zr/sup 95/. These lose their radioactivity within four years and the long- lived radioactivity remains with the ash, which is stored. (AEC) V) nitrate. (P.C.H.)

PLATINUM HEXAFLUORIDE AND METHOD OF FLUORINATING PLUTONIUM CONTAINING MIXTURES THERE-WITH

DOEpatents

Malm, J.G.; Weinstock, B.; Claassen, H.H.

1959-07-01

The preparation of platinum hexafluoride and its use as a fluorinating agent in a process for separating plutonium from fission products is presented. According to the invention, platinum is reacted with fluorine gas at from 900 to 1100 deg C to form platinum hexafluoride. The platinum hexafluoride is then contacted with the plutonium containing mixture at room temperature to form plutonium hexafluoride which is more volatile than the fission products fluorides and therefore can be isolated by distillation.

Protective isolation in single-bed rooms: studies in a modified hospital ward

PubMed Central

Ayliffe, G. A. J.; Collins, B. J.; Lowbury, E. J. L.; Wall, Mary

1971-01-01

Studies were made in a modified hospital ward containing 19 beds, 14 of them in the open ward, one in a window-ventilated side-room, two in rooms with partial-recirculation ventilators giving 7-10 air changes per hour, and two in self-contained isolation suites with plenum ventilation (20 air changes per hour), ultra-violet (UV) barriers at doorways and airlocks. Preliminary tests with aerosols of tracer bacteria showed that few bacteria entered the plenum or recirculation-ventilated rooms. Bacteria released inside mechanically ventilated cubicles escaped into the corridor, but this transfer was reduced by the presence of an airlock. UV barriers at the entrance to the airlock and the cubicle reduced the transfer of bacteria from cubicle to corridor. During a period of 4 years while the ward was in use for surgical and gynaecological patients, the incidence of post-operative sepsis and colonization of wounds by multiple-resistant Staphylococcus aureus was lower (though not significantly lower) in the plenum-ventilated rooms than in the open ward, the recirculator-ventilated cubicles and the window-ventilated cubicles. Nasal acquisition of multiple-resistant Staph. aureus was significantly less common in the plenum-ventilated than in the recirculator-ventilated cubicles and in the other areas. Mean counts of bacteria on settle-plates were significantly lower in the plenum-ventilated cubicles than in the other areas; mean settle-plate counts in the recirculator-ventilated cubicles were significantly lower than in the open ward and in the window-ventilated side-room; similar results were shown by slit-sampling of air. Mean settle-plate counts were significantly lower in all areas when the ward was occupied by female patients. Staph. aureus was rarely carried by air from plenum-ventilated or other cubicles to the open ward, or from the open ward to the cubicles; though staphylococci were transferred from one floor area to another, they did not appear to be redispersed into the air in sufficient numbers to infect the patients. Ultra-violet irradiation caused a significant reduction in the total and staphylococcal counts from the floors of airlocks, and a significant reduction of total counts in the air. PMID:5289715

Numerical Simulation of Metallic Uranium Sintering

NASA Astrophysics Data System (ADS)

Berry, Bruce

Conventional ceramic oxide nuclear fuels are limited in their thermal and life-cycle properties. The desire to operate at higher burnups as is required by current utility economics has proven a formidable challenge for oxide fuel designs. Metallic formulations have superior thermal performance but are plagued by volumetric swelling due to fission gas buildup. In this study, we consider a number of specific microstructure configurations that have been experimentally shown to exhibit considerable resistance to porosity loss. Specifically, a void sizing that is bimodally distributed was shown to resist early pore loss and could provide collection sites for fission gas buildup. We employ the phase field model of Cahn and Hilliard, solved via the finite element method using the open source Multi-User Object Oriented Simulation Environment (MOOSE) developed by INL.

Dry method for recycling iodine-loaded silver zeolite

DOEpatents

Thomas, Thomas R.; Staples, Bruce A.; Murphy, Llewellyn P.

1978-05-09

Fission product iodine is removed from a waste gas stream and stored by passing the gas stream through a bed of silver-exchanged zeolite until the zeolite is loaded with iodine, passing dry hydrogen gas through the bed to remove the iodine and regenerate the bed, and passing the hydrogen stream containing the hydrogen iodide thus formed through a lead-exchanged zeolite which adsorbs the radioactive iodine from the gas stream and permanently storing the lead-exchanged zeolite loaded with radioactive iodine.

Interpretation and modelling of fission product Ba and Mo releases from fuel

NASA Astrophysics Data System (ADS)

Brillant, G.

2010-02-01

The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

Scoping studies of vapor behavior during a severe accident in a metal-fueling reactor

NASA Astrophysics Data System (ADS)

Spencer, B. W.; Marchaterre, J. F.

1985-04-01

The consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel were examined. The principal gas and vapor species released are shown to be Xe, Cs, and bond sodium contained within the fuel porosity. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core. If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the ability of vapor expansion to perform appreciable work on the system and the ability of an expanding vapor bubble to transport fuel and fission produce species to the cover gas region where they may be released to the containment are largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool.

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation

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

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

2016-11-01

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation andmore » evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.« less

Solvent vapor collector

DOEpatents

Ellison, Kenneth; Whike, Alan S.

1979-01-30

A solvent vapor collector is mounted on the upstream inlet end of an oven having a gas-circulating means and intended for curing a coating applied to a strip sheet metal at a coating station. The strip sheet metal may be hot and solvent vapors are evaporated at the coating station and from the strip as it passes from the coating station to the oven. Upper and lower plenums within a housing of the collector are supplied with oven gases or air from the gas-circulating means and such gases or air are discharged within the collector obliquely in a downstream direction against the strip passing through that collector to establish downstream gas flows along the top and under surfaces of the strip so as, in turn, to induct solvent vapors into the collector at the coating station. A telescopic multi-piece shroud is usefully provided on the housing for movement between an extended position in which it overlies the coating station to collect solvent vapors released thereat and a retracted position permitting ready cleaning and adjustment of that coating station.

Massive Gas Injection Valve Development for NSTX-U

DOE Data Explorer

Raman, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Plunkett, G. J. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Way, W.-S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2016-05-01

NSTX-U research will offer new insight by studying gas assimilation efficiencies for MGI injection from different poloidal locations using identical gas injection systems. In support of this activity, an electromagnetic MGI valve has been built and tested. The valve operates by repelling two conductive disks due to eddy currents induced on them by a rapidly changing magnetic field created by a pancake disk solenoid positioned beneath the circular disk attached to a piston. The current is driven in opposite directions in the two solenoids, which creates a cancelling torque when the valve is operated in an ambient magnetic field, as would be required in a tokamak installation. The valve does not use ferromagnetic materials. Results from the operation of the valve, including tests conducted in 1 T external magnetic fields, are described. The pressure rise in the test chamber is measured directly using a fast time response baratron gauge. At a plenum pressure of just 1.38 MPa (~200 psig), the valve injects 27 Pa.m^3 (~200 Torr.L) of nitrogen with a pressure rise time of 3 ms.

Gaseous and particulate emissions from a DC arc melter.

PubMed

Overcamp, Thomas J; Speer, Matthew P; Griner, Stewart J; Cash, Douglas M

2003-01-01

Tests treating soils contaminated with metal compounds and radionuclide surrogates were conducted in a DC arc melter. The soil melted, and glassy or ceramic waste forms with a separate metal phase were produced. Tests were run in the melter plenum with either air or N2 purge gases. In addition to nitrogen, the primary emissions of gases were CO2, CO, oxygen, methane, and oxides of nitrogen (NO(x)). Although the gas flow through the melter was low, the particulate concentrations ranged from 32 to 145 g/m3. Cerium, a nonradioactive surrogate for plutonium and uranium, was not enriched in the particulate matter (PM). The PM was enriched in cesium and highly enriched in lead.

Apparatus for measuring the decontamination factor of a multiple filter air-cleaning system

DOEpatents

Ortiz, John P.

1986-01-01

An apparatus for measuring the overall decontamination factor of first and second filters located in a plenum. The first filter separates the plenum's upstream and intermediate chambers. The second filter separates the plenum's intermediate and downstream chambers. The apparatus comprises an aerosol generator that generates a challenge aerosol. An upstream collector collects unfiltered aerosol which is piped to first and second dilution stages and then to a laser aerosol spectrometer. An intermediate collector collects challenge aerosol that penetrates the first filter. The filtered aerosol is piped to the first dilution stage, diluted, and then piped to the laser aerosol spectrometer which detects single particles. A downstream collector collects challenge aerosol that penetrates both filters. The twice-filtered aerosol is piped to the aerosol spectrometer. A pump and several valves control the movement of aerosol within the apparatus.

Apparatus for measuring the decontamination factor of a multiple filter air-cleaning system

DOEpatents

Ortiz, J.P.

1985-07-03

An apparatus for measuring the overall decontamination factors of first and second filters located in a plenum. The first filter separates the plenum's upstream and intermediate chambers. The second filter separates the plenum's intermediate and downstream chambers. The apparatus comprises an aerosol generator that generates a challenge aerosol. An upstream collector collects unfiltered aerosol which is piped to first and second dilution stages and then to a laser aerosol spectrometer. An intermediate collector collects challenge aerosol that penetrates the first filter. The filtered aerosol is piped to the first dilution stage, diluted, and then piped to the laser aerosol spectrometer which detects single particles. A downstream collector collects challenge aerosol that penetrates both filters. The twice-filtered aerosol is piped to the aerosol spectrometer. A pump and several valves control the movement of aerosol within the apparatus.

Modeling study of deposition locations in the 291-Z plenum

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

Mahoney, L.A.; Glissmeyer, J.A.

The TEMPEST (Trent and Eyler 1991) and PART5 computer codes were used to predict the probable locations of particle deposition in the suction-side plenum of the 291-Z building in the 200 Area of the Hanford Site, the exhaust fan building for the 234-5Z, 236-Z, and 232-Z buildings in the 200 Area of the Hanford Site. The Tempest code provided velocity fields for the airflow through the plenum. These velocity fields were then used with TEMPEST to provide modeling of near-floor particle concentrations without particle sticking (100% resuspension). The same velocity fields were also used with PART5 to provide modeling ofmore » particle deposition with sticking (0% resuspension). Some of the parameters whose importance was tested were particle size, point of injection and exhaust fan configuration.« less

Development work for a borax internal core-catcher for a gas-cooled fast reactor

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

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

The Binary Fission Model for the Formation of the Pluto system

NASA Astrophysics Data System (ADS)

Prentice, Andrew

2016-10-01

The ratio F of the mass of Pluto (P) to Charon (C), viz. F ≈ 8:1, is the largest ratio of any planet-satellite pair in the solar system. Another measure of the PC binary is its normalized angular momentum density J (see McKinnon 1989). Analysis of astrometric data (Brozovic et al 2015) acquired before the New Horizons (NH) arrival at Pluto and new measurements made by NH (Stern et al 2015) show that J = 0.39. Yet these F & J values are ones expected if the PC binary had formed by the rotational fission of a single liquid mass (Darwin 1902; Lyttleton 1953). At first glance, therefore, the fission model seems to be a viable model for the formation of the Pluto system. In fact, Prentice (1993 Aust J Astron 5 111) had used this model to successfully predict the existence of several moons orbiting beyond Charon, before their discovery in 2005-2012. The main problem with the fission model is that the observed mean density of Charon, namely 1.70 g/cm3, greatly exceeds that of water ice. Charon thus could not have once been a globe of pure water. Here I review the fission model within the framework of the modern Laplacian theory of solar system origin (Prentice 1978 Moon Planets 19 341; 2006 PASA 23 1) and the NH results. I assume that Pluto and Charon were initially a single object (proto-Pluto [p-P]) which had condensed within the same gas ring shed by the proto-solar cloud at orbital distance ~43 AU, where the Kuiper belt was born. The temperature of this gas ring is 26 K and the mean orbit pressure is 1.3 × 10-9 bar. After the gas ring is shed, chemical condensation takes place. The bulk chemical composition of the condensate is anhydrous rock (mass fraction 0.5255), graphite (0.0163), water ice (0.1858), CO2 ice (0.2211) and methane ice (0.0513). Next I assume that melting of the ices in p-P takes place through the decay of short-lived radioactive nuclides, thus causing internal segregation of the rock & graphite. Settling of heavy grains to the centre lowers the MOI of p-P, so triggering rotational disruption. Pluto's moons would then form from liquid water and liquid CO2, as well as entrained rock-graphite grains. Charon's mean density implies that the rock-graphite mass fraction of the fissioned mass was ˜0.41.

Growth of the interaction layer around fuel particles in dispersion fuel

NASA Astrophysics Data System (ADS)

Olander, D.

2009-01-01

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.

Gas emission from the UO2 samples, containing fission products and burnable absorber

NASA Astrophysics Data System (ADS)

Kopytin, V. P.; Baranov, V. G.; Burlakova, M. A.; Tenishev, A. V.; Kuzmin, R. S.; Pokrovskiy, S. A.; Mikhalchik, V. V.

2016-04-01

The process gas released from the fuel pellets of uranium fuel during fuel burn-up reduces the thermal conductivity of the rod-shell gap, enhances hydrogen embrittlement of the cladding material, causes it's carbonization, as well as transport processes in the fuel. In this study a technique of investigating the thermal desorption of gases from the UO2 fuel material were perfected in the temperature range 300-2000 K for uniform sample heating rate of 15 K/min in vacuum. The characteristic kinetic dependences are acquired for the gas emission from UO2 samples, containing simulators of fission products (SFP) and the burnable neutron absorber (BNA). Depending on the amount of SFP and BNA contained in the sample thermal desorption gas spectra (TDGS) vary. The composition of emitted gas varies, as well as the number of peaks in the TDGS and the peaks shift to higher temperatures. This indicates that introduction of SFPs and BNA alters the sample material structure and cause the creation of so- called traps which have different bonding energies to the gases. The traps can be a grid of dislocations, voids, and contained in the UO2 matrix SFP and BNA. Similar processes will occur in the fuel pellets in the real conditions of the Nuclear Power Plant as well.

Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

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

McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel

2014-04-07

Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development suchmore » that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.« less

Detector Calibration to Spontaneous Fission for the Study of Superheavy Elements Using Gas-Filled Recoil Ion Separator

NASA Astrophysics Data System (ADS)

Takeyama, Mirei; Kaji, Daiya; Morimoto, Kouji; Wakabayashi, Yasuo; Tokanai, Fuyuki; Morita, Kosuke

Detector response to spontaneous fission (SF) of heavy nuclides produced in the 206Pb(48Ca,2n)252No reaction was investigated using a gas-filled recoil ion separator (GARIS). Kinetic energy distributions of the SF originating from 252No were observed by tuning implantation depth of evaporation residue (ER) to the detector. The focal plane detector used in the GARIS experiments was well calibrated by comparing with the known total kinetic energy (TKE) of SF due to 252No. The correction value for the TKE calculation was deduced as a function of the implantation depth of 252No to the detector. Furthermore, we have investigated the results by comparing with those obtained by a computer simulation using the particle and heavy ion transport code system (PHITS).

CANNED SLUG

DOEpatents

Burton, M.

1959-02-17

Fuel elements of the type comprised of a core of fissionable material enclosed in a jacket of nonfissionable, corrosion resistant material are presented. In this invention the fissionable core is shorter than the jacket member, to provide a void chamber in one end of the assembled element. The fissionable material is separated from the chamber by an inwardly extending portion of the jacket member containing a gas permeable wafer centrally disposed therein. The outer end of the chamber is closed bv the end of the jacket which has a rupture disk centrally disposed therein. Gases formed by the irradiation of the fissionable material pass through the porous wafer into the chanmber thereby causing a gradual increase in the pressure in the chamber. The rupture disk is designed to fail at a lower pressure than that which would rupture the jacket. Upon rupture of the disk, the gases in the chamber escape into the coolant channel and coolant enters the chamber but is prevented from coming into contact with the fissionable material by the action of the gases under pressure passing outwardly through the wafer. The ruptured fuel element may be readily detected by monitoring the reactor coolant system.

Formation of Pluto's moons: the fission hypothesis revisited

NASA Astrophysics Data System (ADS)

Prentice, A. J.

2015-12-01

I re-examine the fission hypothesis for the formation of Pluto's moons within the framework of a gas ring model for the origin of the solar system (Prentice 1978 Moon Planets 19 341; 2015 LPSC, abs. 2664). It is supposed that the planetary system condensed from a concentric family of orbiting gas rings. These were cast off by the proto-solar cloud (PSC) as a means for disposing of excess spin angular momentum during gravitational contraction. If contraction is homologous, the mean orbital radii R(n) (n = 0,1,2,3,..) of the rings form a nearly geometric sequence. The temperatures T(n) of the rings scale roughly as T(n) = A/R(n) and the gas pressures p(n) on the gas ring mean orbits scale as p(n) = B/R(n)^4. The constants A & B are chosen so that (1) the geometric mean of the ratio R(n+1)/R(n) of successive gas ring radii from Jupiter to Mercury matches the observed mean ratio of planetary distances and (2) that the metal mass fraction at Mercury's orbit, namely 0.70, yields a planet whose mean density equals the observed value (Prentice 2008, LPSC abs. 1945.pdf). I assume that proto-Pluto (PPO) condensed within the n = 0 gas ring shed by the PSC at the orbit of Quaoar (43.2 AU). Here T(0) = 26.3 K and p(0) = 1.3 x 10^(-9) bar. The condensate consists of anhydrous rock (mass fraction 0.5255), graphite (0.0163), water ice (0.1858), dry ice (0.2211), and methane ice (0.0513). The RTP rock density is 3.662 g/cc. I assume that melting of the ices in the PPO took place through the decay of short-lived radioactive nuclides, causing internal segregation of rock & graphite. If rotational fission did occur and Pluto's moons formed from ejected liquid water and CO2, we get a Charon mean density of 1.24 g/cc. This is much lower than the observed value. Perhaps some of the rock and graphite became entrained in the fissioned liquid, so yielding a dense core for Charon of mass fraction ~0.4? In any event, the surfaces of all of the moons should have initially been football-shaped, very smooth and consist solely of water ice. As there is no outward migration of the major planets in the gas ring model, the risk of impact bombardment is minimal. Most likely, subsequent tidal action between Pluto and Charon produced the chasms that girdle the equator of Charon (Barr & Collins 2015). I predict that New Horizons will detect dry ice in those parts of Hydra that have been gouged by impacts.

Effects of bleed-hole geometry and plenum pressure on three-dimensional shock-wave/boundary-layer/bleed interactions

NASA Technical Reports Server (NTRS)

Chyu, Wei J.; Rimlinger, Mark J.; Shih, Tom I.-P.

1993-01-01

A numerical study was performed to investigate 3D shock-wave/boundary-layer interactions on a flat plate with bleed through one or more circular holes that vent into a plenum. This study was focused on how bleed-hole geometry and pressure ratio across bleed holes affect the bleed rate and the physics of the flow in the vicinity of the holes. The aspects of the bleed-hole geometry investigated include angle of bleed hole and the number of bleed holes. The plenum/freestream pressure ratios investigated range from 0.3 to 1.7. This study is based on the ensemble-averaged, 'full compressible' Navier-Stokes (N-S) equations closed by the Baldwin-Lomax algebraic turbulence model. Solutions to the ensemble-averaged N-S equations were obtained by an implicit finite-volume method using the partially-split, two-factored algorithm of Steger on an overlapping Chimera grid.

Internal baffling for fuel injector

DOEpatents

Johnson, Thomas Edward; Lacy, Benjamin; Stevenson, Christian

2014-08-05

A fuel injector includes a fuel delivery tube; a plurality of pre-mixing tubes, each pre-mixing tube comprising at least one fuel injection hole; an upstream tube support plate that supports upstream ends of the plurality of pre-mixing tubes; a downstream tube support plate that supports downstream ends of the plurality of pre-mixing tubes; an outer wall connecting the upstream tube support plate and the downstream tube support plate and defining a plenum therewith; and a baffle provided in the plenum. The baffle includes a radial portion. A fuel delivered in the upstream direction by the fuel delivery tube is directed radially outwardly in the plenum between the radial portion of the baffle and the downstream tube support plate, then in the downstream direction around an outer edge portion of the radial portion, and then radially inwardly between the radial portion and the upstream tube support plate.

Multi-tube thermal fuse for nozzle protection from a flame holding or flashback event

DOEpatents

Lacy, Benjamin Paul; Davis, Jr., Lewis Berkley; Johnson, Thomas Edward; York, William David

2012-07-03

A protection system for a pre-mixing apparatus for a turbine engine, includes: a main body having an inlet portion, an outlet portion and an exterior wall that collectively establish a fuel delivery plenum; and a plurality of fuel mixing tubes that extend through at least a portion of the fuel delivery plenum, each of the plurality of fuel mixing tubes including at least one fuel feed opening fluidly connected to the fuel delivery plenum; at least one thermal fuse disposed on an exterior surface of at least one tube, the at least one thermal fuse including a material that will melt upon ignition of fuel within the at least one tube and cause a diversion of fuel from the fuel feed opening to at least one bypass opening. A method and a turbine engine in accordance with the protection system are also provided.

Production of beta-gamma coincidence spectra of individual radioxenon isotopes for improved analysis of nuclear explosion monitoring data

NASA Astrophysics Data System (ADS)

Haas, Derek Anderson

Radioactive xenon gas is a fission product released in the detonation of nuclear devices that can be detected in atmospheric samples far from the detonation site. In order to improve the capabilities of radioxenon detection systems, this work produces beta-gamma coincidence spectra of individual isotopes of radioxenon. Previous methods of radioxenon production consisted of the removal of mixed isotope samples of radioxenon gas released from fission of contained fissile materials such as 235U. In order to produce individual samples of the gas, isotopically enriched stable xenon gas is irradiated with neutrons. The detection of the individual isotopes is also modeled using Monte Carlo simulations to produce spectra. The experiment shows that samples of 131mXe, 133 Xe, and 135Xe with a purity greater than 99% can be produced, and that a sample of 133mXe can be produced with a relatively low amount of 133Xe background. These spectra are compared to models and used as essential library data for the Spectral Deconvolution Analysis Tool (SDAT) to analyze atmospheric samples of radioxenon for evidence of nuclear events.

Laser-assisted isotope separation of tritium

DOEpatents

Herman, Irving P.; Marling, Jack B.

1983-01-01

Methods for laser-assisted isotope separation of tritium, using infrared multiple photon dissociation of tritium-bearing products in the gas phase. One such process involves the steps of (1) catalytic exchange of a deuterium-bearing molecule XYD with tritiated water DTO from sources such as a heavy water fission reactor, to produce the tritium-bearing working molecules XYT and (2) photoselective dissociation of XYT to form a tritium-rich product. By an analogous procedure, tritium is separated from tritium-bearing materials that contain predominately hydrogen such as a light water coolant from fission or fusion reactors.

Payload dose rate from direct beam radiation and exhaust gas fission products. [for nuclear engine for rocket vehicles

NASA Technical Reports Server (NTRS)

Capo, M. A.; Mickle, R.

1975-01-01

A study was made to determine the dose rate at the payload position in the NERVA System (1) due to direct beam radiation and (2) due to the possible effect of fission products contained in the exhaust gases for various amounts of hydrogen propellant in the tank. Results indicate that the gamma radiation is more significant than the neutron flux. Under different assumptions the gamma contribution from the exhaust gases was 10 to 25 percent of total gamma flux.

Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 4: Energy from fossil fuels

NASA Technical Reports Server (NTRS)

Williams, J. R.

1974-01-01

The conversion of fossil-fired power plants now burning oil or gas to burn coal is discussed along with the relaxation of air quality standards and the development of coal gasification processes to insure a continued supply of gas from coal. The location of oil fields, refining areas, natural gas fields, and pipelines in the U.S. is shown. The technologies of modern fossil-fired boilers and gas turbines are defined along with the new technologies of fluid-bed boilers and MHD generators.

An experimental investigation of 235 sub UF sub 6 fission produced plasmas. [gas handling system for use with nuclear pumped laser experiments

NASA Technical Reports Server (NTRS)

Miley, G. H.

1981-01-01

A gas handling system capable of use with uranium fluoride was designed and constructed for use with nuclear pumped laser experiments using the TRIGA research reactor. By employing careful design and temperature controls, the UF6 can be first transported into the irradiation chamber, and then, at the conclusion of the experiment, returned to gas cylinders. The design of the system is described. Operating procedures for the UF6 and gas handling systems are included.

Swelling and gas release in oxide fuels during fast temperature transients

NASA Astrophysics Data System (ADS)

Dollins, C. C.; Jursich, M.

1982-05-01

A previously reported intergranular swelling and gas release model for oxide fuels has been modified to predict fission gas behavior during fast temperature transients. Under steady state or slowly varying conditions it has been assumed in the previous model that the pressure caused by the fission gas within the gas bubbles is in equilibrium with the surface tension of the bubbles. During a fast transient, however, net vacancy migration to the bubbles may be insufficient to maintain this equilibrium. In order to ascertain the net vacancy flow, it is necessary to model the point defect behavior in the fuel. Knowing the net flow of vacancies to the bubble and the bubble size, the bubble diffusivity can be determined and the long range migration of the gas out of the fuel can be calculated. The model has also been modified to allow release of all the gas on the grain boundaries during a fast temperature transient. The gas release predicted by the revised model shows good agreement to fast transient gas release data from an EBR-II TREAT H-3 (Transient Reactor Test Facility) test. Agreement has also been obtained between predictions using the model and gas release data obtained by Argonne National Laboratory from out-of-reactor transient heating experiments on irradiated UO 2. It was found necessary to increase the gas bubble diffusivity used in the model by a factor of thirty during the transient to provide agreement between calculations and measurements. Other workers have also found that such an increase is necessary for agreement and attribute the increased diffusivity to yielding at the bubble surface due to the increased pressure.

β(1,3)-Glucanosyl-Transferase Activity Is Essential for Cell Wall Integrity and Viability of Schizosaccharomyces pombe

PubMed Central

de Medina-Redondo, María; Arnáiz-Pita, Yolanda; Clavaud, Cécile; Fontaine, Thierry; del Rey, Francisco; Latgé, Jean Paul; Vázquez de Aldana, Carlos R.

2010-01-01

Background The formation of the cell wall in Schizosaccharomyces pombe requires the coordinated activity of enzymes involved in the biosynthesis and modification of β-glucans. The β(1,3)-glucan synthase complex synthesizes linear β(1,3)-glucans, which remain unorganized until they are cross-linked to other β(1,3)-glucans and other cell wall components. Transferases of the GH72 family play important roles in cell wall assembly and its rearrangement in Saccharomyces cerevisiae and Aspergillus fumigatus. Four genes encoding β(1,3)-glucanosyl-transferases -gas1+, gas2+, gas4+ and gas5+- are present in S. pombe, although their function has not been analyzed. Methodology/Principal Findings Here, we report the characterization of the catalytic activity of gas1p, gas2p and gas5p together with studies directed to understand their function during vegetative growth. From the functional point of view, gas1p is essential for cell integrity and viability during vegetative growth, since gas1Δ mutants can only grow in osmotically supported media, while gas2p and gas5p play a minor role in cell wall construction. From the biochemical point of view, all of them display β(1,3)-glucanosyl-transferase activity, although they differ in their specificity for substrate length, cleavage point and product size. In light of all the above, together with the differences in expression profiles during the life cycle, the S. pombe GH72 proteins may accomplish complementary, non-overlapping functions in fission yeast. Conclusions/Significance We conclude that β(1,3)-glucanosyl-transferase activity is essential for viability in fission yeast, being required to maintain cell integrity during vegetative growth. PMID:21124977

Correlation of recent fission product release data

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

Kress, T.S.; Lorenz, R.A.; Nakamura, T.

For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS butmore » which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab.« less

Phase 1 Space Fission Propulsion System Design Considerations

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Carter, Robert; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Rodgers, Stephen L. (Technical Monitor)

2001-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems operating at 80 kWe or above could enhance or enable numerous robotic outer solar system missions of interest. At these power levels it is possible to develop safe, affordable systems that meet mission performance requirements. In selecting the system design to pursue, seven evaluation criteria were identified: safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of three potential concepts was performed: an SP-100 based pumped liquid lithium system, a direct gas cooled system, and a heatpipe cooled system. For power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the heatpipe system has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the heatpipe approach. Successful development and utilization of a "Phase 1" fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system.

Phase 1 space fission propulsion system design considerations

NASA Astrophysics Data System (ADS)

Houts, Mike; van Dyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems operating at 80 kWe or above could enhance or enable numerous robotic outer solar system missions of interest. At these power levels it is possible to develop safe, affordable systems that meet mission performance requirements. In selecting the system design to pursue, seven evaluation criteria were identified: safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of three potential concepts was performed: an SP-100 based pumped liquid lithium system, a direct gas cooled system, and a heatpipe cooled system. For power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the heatpipe system has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the heatpipe approach. Successful development and utilization of a ``Phase 1'' fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system. .

Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

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

Jason M. Harp; Paul A. Demkowicz

2014-10-01

In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10 -4 to 10 -5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materialsmore » is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.« less

The role of off-line mass spectrometry in nuclear fission.

PubMed

De Laeter, J R

1996-01-01

The role of mass spectrometry in nuclear fission has been invaluable since 1940, when A. O. C. Nier separated microgram quantities of (235) U from (238) U, using a gas source mass spectrometer. This experiment enabled the fissionable nature of (235) U to be established. During the Manhattan Project, the mass spectrometer was used to measure the isotope abundances of uranium after processing in various separation systems, in monitoring the composition of the gaseous products in the Oak Ridge Diffusion Plant, and as a helium leak detector. Following the construction of the first reactor at the University of Chicago, it was necessary to unravel the nuclear systematics of the various fission products produced in the fission process. Off-line mass spectrometry was able to identify stable and long-lived isotopes produced in fission, but more importantly, was used in numerous studies of the distribution of mass of the cumulative fission yields. Improvements in sensitivity enabled off-line mass spectrometric studies to identify fine structure in the mass-yield curve and, hence, demonstrate the importance of shell structure in nuclear fission. Solid-source mass spectrometry was also able to measure the cumulative fission yields in the valley of symmetry in the mass-yield curve, and enabled spontaneous fission yields to be quantified. Apart from the accurate measurement of abundances, the stable isotope mass spectrometric technique has been invaluable in establishing absolute cumulative fission yields for many isotopes making up the mass-yield distribution curve for a variety of fissile nuclides. Extensive mass spectrometric studies of noble gases in primitive meteorites revealed the presence of fission products from the now extinct nuclide (244) Pu, and have eliminated the possibility of fission products from a super-heavy nuclide contributing to isotopic anomalies in meteoritic material. Numerous mass spectrometric studies of the isotopic and elemental abundances of samples from the Oklo Natural Reactor have enabled the nuclear parameters of the various reactor zones to be calculated, and the mobility/retentivity of a number of elements to be established in the reactor zones and the surrounding rocks. These isotopic studies have given valuable information on the geochemical behavior of natural geological repositories for radioactive waste containment. © 1997 John Wiley & Sons, Inc. Copyright © 1997 John Wiley & Sons, Inc.

Toxicities of RDX or TNT Freshly Amended or Weathered-and-Aged in Five Natural Soils to the Collembolan Folsomia candida

DTIC Science & Technology

2013-05-01

Biodegradation of Nitroaromatic Compounds; Spain, J.C., Ed.; Plenum Press: New York , 1995; pp 69–85. Renoux, A.Y.; Sarrazin, M.; Hawari, J.; Sunahara...in Contaminated Soil. In Biodegradation of Nitroaromatic Compounds. Spain, J.C., Ed.; Plenum Press: New York , 1995; pp 1–18. Robidoux, P.Y...CANDIDA 1. INTRODUCTION Many sites associated with military operations that involve munition manufacturing, disposal , testing, and training

Rarefied gas electro jet (RGEJ) micro-thruster for space propulsion

NASA Astrophysics Data System (ADS)

Blanco, Ariel; Roy, Subrata

2017-11-01

This article numerically investigates a micro-thruster for small satellites which utilizes plasma actuators to heat and accelerate the flow in a micro-channel with rarefied gas in the slip flow regime. The inlet plenum condition is considered at 1 Torr with flow discharging to near vacuum conditions (<0.05 Torr). The Knudsen numbers at the inlet and exit planes are ~0.01 and ~0.1, respectively. Although several studies have been performed in micro-hallow cathode discharges at constant pressure, to our knowledge, an integrated study of the glow discharge physics and resulting fluid flow of a plasma thruster under these low pressure and low Knudsen number conditions is yet to be reported. Numerical simulations of the charge distribution due to gas ionization processes and the resulting rarefied gas flow are performed using an in-house code. The mass flow rate, thrust, specific impulse, power consumption and the thrust effectiveness of the thruster are predicted based on these results. The ionized gas is modelled using local mean energy approximation. An electrically induced body force and a thermal heating source are calculated based on the space separated charge distribution and the ion Joule heating, respectively. The rarefied gas flow with these electric force and heating source is modelled using density-based compressible flow equations with slip flow boundary conditions. The results show that a significant improvement of specific impulse can be achieved over highly optimized cold gas thrusters using the same propellant.

Studies of Fission Fragment Rocket Engine Propelled Spacecraft

NASA Technical Reports Server (NTRS)

Werka, Robert O.; Clark, Rodney; Sheldon, Rob; Percy, Thomas K.

2014-01-01

The NASA Office of Chief Technologist has funded from FY11 through FY14 successive studies of the physics, design, and spacecraft integration of a Fission Fragment Rocket Engine (FFRE) that directly converts the momentum of fission fragments continuously into spacecraft momentum at a theoretical specific impulse above one million seconds. While others have promised future propulsion advances if only you have the patience, the FFRE requires no waiting, no advances in physics and no advances in manufacturing processes. Such an engine unequivocally can create a new era of space exploration that can change spacecraft operation. The NIAC (NASA Institute for Advanced Concepts) Program Phase 1 study of FY11 first investigated how the revolutionary FFRE technology could be integrated into an advanced spacecraft. The FFRE combines existent technologies of low density fissioning dust trapped electrostatically and high field strength superconducting magnets for beam management. By organizing the nuclear core material to permit sufficient mean free path for escape of the fission fragments and by collimating the beam, this study showed the FFRE could convert nuclear power to thrust directly and efficiently at a delivered specific impulse of 527,000 seconds. The FY13 study showed that, without increasing the reactor power, adding a neutral gas to the fission fragment beam significantly increased the FFRE thrust through in a manner analogous to a jet engine afterburner. This frictional interaction of gas and beam resulted in an engine that continuously produced 1000 pound force of thrust at a delivered impulse of 32,000 seconds, thereby reducing the currently studied DRM 5 round trip mission to Mars from 3 years to 260 days. By decreasing the gas addition, this same engine can be tailored for much lower thrust at much higher impulse to match missions to more distant destinations. These studies created host spacecraft concepts configured for manned round trip journeys. While the vehicles are very large, they are primarily made up of a habitat payload on one end, the engine on the opposite end and a connecting spine containing radiator acreage needed to reject the heat of this powerful, but inefficient engine. These studies concluded that the engine and spacecraft are within today's technology, could be built, tested, launched on several SLS launchers, integrated, checked out, maintained at an in-space LEO base, and operated for decades just as Caribbean cruise ships operate today. The nuclear issues were found to be far less daunting that [than for] current nuclear engines. The FFRE produces very small amounts of radioactive efflux compared to their impulse, easily contained in an evacuated "bore-hole" test site. The engine poses no launch risk since it is simply a structure containing no fissionable material. The nuclear fuel is carried to orbit in containers highly crash-proofed for launch accidents from which it, in a liquid medium, is injected into the FFRE. The radioactive exhaust, with a velocity above 300 kilometers per second rapidly leaves the solar system.

A Numerical Investigation of the Startup Transient in a Wave Rotor

NASA Technical Reports Server (NTRS)

Paxson, Daniel E.

1996-01-01

The startup process is investigated for a hypothetical four-port wave rotor, envisioned as a topping cycle for a small gas turbine engine. The investigation is conducted numerically using a multi-passage, one-dimensional CFD-based wave rotor simulation in combination with lumped volume models for the combustor, exhaust valve plenum, and rotor center cavity components. The simulation is described and several startup transients are presented which illustrate potential difficulties for the specific cycle design investigated. In particular it is observed that, prior to combustor light-off, or just after, the flow through the combustor loop is reversed from the design direction. The phenomenon is demonstrated and several possible modifications techniques are discussed which avoid or overcome the problem.

Exciton fission in monolayer transition metal dichalcogenide semiconductors.

PubMed

Steinhoff, A; Florian, M; Rösner, M; Schönhoff, G; Wehling, T O; Jahnke, F

2017-10-27

When electron-hole pairs are excited in a semiconductor, it is a priori not clear if they form a plasma of unbound fermionic particles or a gas of composite bosons called excitons. Usually, the exciton phase is associated with low temperatures. In atomically thin transition metal dichalcogenide semiconductors, excitons are particularly important even at room temperature due to strong Coulomb interaction and a large exciton density of states. Using state-of-the-art many-body theory, we show that the thermodynamic fission-fusion balance of excitons and electron-hole plasma can be efficiently tuned via the dielectric environment as well as charge carrier doping. We propose the observation of these effects by studying exciton satellites in photoemission and tunneling spectroscopy, which present direct solid-state counterparts of high-energy collider experiments on the induced fission of composite particles.

Development of a full-length external-fuel thermionic converter for in-pile testing.

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

Description of an external-fuel thermionic converter which utilizes a thoriated-tungsten fuel-emitter body. Performance in out-of-pile tests was comparable to that of an arc-cast tungsten emitter body, with 400-eW output power (about 5 W/sq cm) at 10.8% efficiency. Maximum fuel clad temperature averaged from 1650 to 1700 C during the 300-hour test. This converter has been processed for in-pile testing. The various processing steps, including the installation of six emitter thermocouples, encapsulation in the secondary container, and joining to the fission-gas collection system, are described in detail. In addition to the converter assembly, a doubly contained fission gas collection assembly with radiation-hardened differential pressure transducers was fabricated. The experiment support plate required for the in-pile test, containing electrically insulated instrumentation feedthroughs and coolant line feedthroughs to the vacuum test chamber, was also fabricated.

Parametric analyses of planned flowing uranium hexafluoride critical experiments

NASA Technical Reports Server (NTRS)

Rodgers, R. J.; Latham, T. S.

1976-01-01

Analytical investigations were conducted to determine preliminary design and operating characteristics of flowing uranium hexafluoride (UF6) gaseous nuclear reactor experiments in which a hybrid core configuration comprised of UF6 gas and a region of solid fuel will be employed. The investigations are part of a planned program to perform a series of experiments of increasing performance, culminating in an approximately 5 MW fissioning uranium plasma experiment. A preliminary design is described for an argon buffer gas confined, UF6 flow loop system for future use in flowing critical experiments. Initial calculations to estimate the operating characteristics of the gaseous fissioning UF6 in a confined flow test at a pressure of 4 atm, indicate temperature increases of approximately 100 and 1000 K in the UF6 may be obtained for total test power levels of 100 kW and 1 MW for test times of 320 and 32 sec, respectively.

A semi-empirical model for the formation and depletion of the high burnup structure in UO 2

DOE PAGES

Pizzocri, D.; Cappia, F.; Luzzi, L.; ...

2017-01-31

In the rim zone of UO 2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. To this end, we per-formed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Moreover, based on these new experimental data, we assume an exponential reduction of the average grain size withmore » local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.« less

Effects of Irradiation on the Microstructure of U-7Mo Dispersion Fuel with Al-2Si Matrix

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

Dennis D. Keiser, Jr.; Jan-Fong Jue; Adam B. Robinson

2012-06-01

The Reduced Enrichment for Research and Test Reactor program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt% Si added to the matrix, fuel plates were tested to medium burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fissionmore » rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, high fission rate) was performed in the RERTR-9A, RERTR-9B and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth of the fuel/matrix interaction layer (FMI), which was present in the samples to some degree after fabrication, during irradiation; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation more Si diffuses from the matrix to the FMI layer/matrix interface, and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.« less

Mechanistic approach for nitride fuel evolution and fission product release under irradiation

NASA Astrophysics Data System (ADS)

Dolgodvorov, A. P.; Ozrin, V. D.

2017-01-01

A model for describing uranium-plutonium mixed nitride fuel pellet burning was developed. Except fission products generating, the model includes impurities of oxygen and carbon. Nitrogen behaviour in nitride fuel was analysed and the nitrogen chemical potential in solid solution with uranium-plutonium nitride was constructed. The chemical program module was tested with the help of thermodynamic equilibrium phase distribution calculation. Results were compared with analogous data in literature, quite good agreement was achieved, especially for uranium sesquinitride, metallic species and some oxides. Calculation of a process of nitride fuel burning was also conducted. Used mechanistic approaches for fission product evolution give the opportunity to find fission gas release fractions and also volumes of intergranular secondary phases. Calculations present that the most massive secondary phases are the oxide and metallic phases. Oxide phase contain approximately 1 % wt of substance over all time of burning with slightly increasing of content. Metallic phase has considerable rising of mass and by the last stage of burning it contains about 0.6 % wt of substance. Intermetallic phase has less increasing rate than metallic phase and include from 0.1 to 0.2 % wt over all time of burning. The highest element fractions of released gaseous fission products correspond to caesium and iodide.

Research Activities in Fission Chamber Modeling in Support of the Nuclear Energy Industry

NASA Astrophysics Data System (ADS)

Jammes, C.; Filliatre, P.; Geslot, B.; Oriol, L.; Berhouet, F.; Villard, J.-F.; Vermeeren, L.

2010-12-01

Fission chambers are widely used in the nuclear industry. As an example, they play a major role in the control of any fission reactor and are thus regarded as a key component for ensuring their safety. They are also employed in the material testing reactors for monitoring irradiations. We have recently started a research program, the objective of which is to improve the performance of those neutron detectors in terms of lifetime, calibration, and online diagnosis. In this paper, we present several studies carried out in order to model the signal delivered by a fission chamber. First, the simulation of the deposit evolution allowed us to select the most appropriate fissile material for a given spectrum and fluence. Second, we studied the impact of the bias voltage and filling gas characteristics on the charge collection time. Finally, the simulation of a pulse signal prior to amplification showed how it is important to have a satisfactory knowledge of the energy for creating ion pairs to accurately assess the signal in current or Campbelling mode.

Publications - GMC 225 | Alaska Division of Geological & Geophysical

Science.gov Websites

following Alaskan arctic White Hills oil and gas wells: ARCO Alaska Inc. Kavik #1 (75' - 9540'); and BP O'Sullivan, P.B., 1994, Apatite fission track data derived from cuttings of the following Alaskan arctic

Detonation duct gas generator demonstration program

NASA Technical Reports Server (NTRS)

Wortman, Andrew; Brinlee, Gayl A.; Othmer, Peter; Whelan, Michael A.

1991-01-01

The feasibility of the generation of detonation waves moving periodically across high speed channel flow is experimentally demonstrated. Such waves are essential to the concept of compressing requirements and increasing the engine pressure compressor with the objective of reducing conventional compressor requirements and increasing the engine thermodynamic efficiency through isochoric energy addition. By generating transient transverse waves, rather than standing waves, shock wave losses are reduced by an order of magnitude. The ultimate objective is to use such detonation ducts downstream of a low pressure gas turbine compressor to produce a high overall pressure ratio thermodynamic cycle. A 4 foot long, 1 inch x 12 inch cross-section, detonation duct was operated in a blow-down mode using compressed air reservoirs. Liquid or vapor propane was injected through injectors or solenoid valves located in the plenum or the duct itself. Detonation waves were generated when the mixture was ignited by a row of spark plugs in the duct wall. Problems with fuel injection and mixing limited the air speeds to about Mach 0.5, frequencies to below 10 Hz, and measured pressure ratios of about 5 to 6. The feasibility of the gas dynamic compression was demonstrated and the critical problem areas were identified.

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors. FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.

Oil flow at the scroll compressor discharge: visualization and CFD simulation

NASA Astrophysics Data System (ADS)

Xu, Jiu; Hrnjak, Pega

2017-08-01

Oil is important to the compressor but has other side effect on the refrigeration system performance. Discharge valves located in the compressor plenum are the gateway for the oil when leaving the compressor and circulate in the system. The space in between: the compressor discharge plenum has the potential to separate the oil mist and reduce the oil circulation ratio (OCR) in the system. In order to provide information for building incorporated separation feature for the oil flow near the compressor discharge, video processing method is used to quantify the oil droplets movement and distribution. Also, CFD discrete phase model gives the numerical approach to study the oil flow inside compressor plenum. Oil droplet size distributions are given by visualization and simulation and the results show a good agreement. The mass balance and spatial distribution are also discussed and compared with experimental results. The verification shows that discrete phase model has the potential to simulate the oil droplet flow inside the compressor.

Time correlated measurements using plastic scintillators with neutron-photon pulse shape discrimination

NASA Astrophysics Data System (ADS)

Richardson, Norman E., IV

Since the beginning of the nuclear age, there has been a strong demand for the development of efficient technologies for the detection of ionizing radiation. According to the United States' Department of Energy, the accurate assessment of fissile materials is essential in achieving the nonproliferation goals of enhancing safety and security of nuclear fuel cycle and nuclear energy facilities. Nuclear materials can be characterized by the measurement of prompt and delayed neutrons and gamma rays emitted in spontaneous or induced fission reactions and neutrons emitted in fission reactions are the distinctive signatures of nuclear materials. Today, the most widely used neutron detection technologies rely on thermal neutron capture reactions using a moderating material to cause the neutron to lose its energy prior to the detection event. This is necessary because as the fission event occurs, neutrons are emitted carrying high amounts of energy, typically on the order of mega electron volts (MeV). These energetic particles are classified as "fast" neutrons. For detecting the thermal neutrons, the Helium-3 (3He) gas-filled counters are arguably the most widely used technology of neutron detection. 3He counters have been the scientific standard for the nuclear engineering community for several decades, and have earned their place as a reliable technique for the detection of neutrons. However, 3He gas-filled counters have several disadvantages. First, gas-filled counters are not rigid and are sensitive to vibrations. Secondly, gas-filled counters are prone to the count rate limitations due to the physical processes of charge multiplication and transport in the gas medium in the electric field. Lastly, 3He gas-filled counters suffer from a supply shortage of the 3He isotope. As it is stated in [3], this shortage is created by the new demand for Helium-3 due to the deployment of neutron detectors at the borders after the 9/11 attack to help secure the nation against smuggled nuclear and radiological material. Moreover, the production of 3He isotope as a byproduct of security programs was drastically decreased. This isotope shortage coupled with the disadvantages of relying on a detector that requires neutron moderation before the detection of fission neutrons, poses a significant challenge in supporting the existing detection systems and the development of future technologies. To address this problem, a reliable and accurate alternative technology to detect neutrons emitted in fissions must be developed. One such alternative technology that shows promise in this application is the use of scintillators based on solid state materials (plastics) which are sensitive to fast neutrons. However, plastic scintillators are also sensitive to photons. Hence, it is necessary to separate the neutron signals from the photon signals, using the pulse shape discrimination (PSD) analysis. The PSD is based on the comparison of the pulse shapes of digitized signal waveforms. This approach allows for the measurement of fast neutrons without the necessity of their moderation. Because the fission spectrum neutrons are mainly fast, methods employing fast neutron detection are applicable for the assay of fissile materials. In addition, the average time of scintillation of the plastic medium is much shorter than those of the gaseous counters, thus allowing scintillation detectors to be used in high count rate environments. Furthermore, the temporal information of the fast neutron detection using multiple sensors enables the time correlation analysis of the fission neutron multiplicity. The study of time correlation measurements of fast neutrons using the array of plastic scintillators is the basis of this work. The array of four plastic scintillator detectors equipped with the digital data acquisition and analysis system was developed. The digital PSD analysis of detector signals "on-the-fly" was implemented for the array. The time coincidence measurement technique using the list mode was employed for two detectors operating on the single time scale. This was necessary as no fission source was available to be used as a fast neutron multiplicity source. The detection technology was tested using isotopic photon sources and a plutonium-beryllium neutron source. It was shown that the system can be effectively used for fast-neutron multiplicity measurements, through a "proof-of-concept" model, enabling a shorter width of the time coincidence window compared to the 3He counters. This result opens prospects to reduce the false coincidence rates in the neutron multiplicity measurements, thus increasing the sensitivity of nuclear material detection.

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

NASA Astrophysics Data System (ADS)

Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

2004-02-01

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

Development of a coolant channel helium and nitrogen gas ratio sensor for a high temperature gas reactor

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

Cadell, S. R.; Woods, B. G.

2012-07-01

To measure the changing gas composition of the coolant during a postulated High Temperature Gas Reactor (HTGR) accident, an instrument is needed. This instrument must be compact enough to measure the ratio of the coolant versus the break gas in an individual coolant channel. This instrument must minimally impact the fluid flow and provide for non-direct signal routing to allow minimal disturbance to adjacent channels. The instrument must have a flexible geometry to allow for the measurement of larger volumes such as in the upper or lower plenum of a HTGR. The instrument must be capable of accurately functioning throughmore » the full operating temperature and pressure of a HTGR. This instrument is not commercially available, but a literature survey has shown that building off of the present work on Capacitance Sensors and Cross-Capacitors will provide a basis for the development of the desired instrument. One difficulty in developing and instrument to operate at HTGR temperatures is acquiring an electrical conductor that will not melt at 1600 deg. C. This requirement limits the material selection to high temperature ceramics, graphite, and exotic metals. An additional concern for the instrument is properly accounting for the thermal expansion of both the sensing components and the gas being measured. This work covers the basic instrument overview with a thorough discussion of the associated uncertainty in making these measurements. (authors)« less

Milestone report: The simulation of radiation driven gas diffusion in UO 2 at low temperature

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

Cooper, Michael William; Kuganathan, Navaratnarajah; Burr, Patrick A

2016-10-24

Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. This is an important process for nuclear reactor performance as it affects fission gas release, particularly from the periphery of the pellet where such temperatures are normal. Here we present a molecular dynamics study of Xe and Kr diffusion due to irradiation. Thermal spikes and cascades have been used to study the electronic stopping and ballistic phases of damage, respectively. Our results predict that O and Kr exhibit the greatest diffusivity and U the least, while Xemore » lies in between. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Preliminary thermal spike calculations indicate that the electronic stopping phase generates greater fission gas displacement than the ballistic phase, although further calculation must be carried out to confirm this. A good description of the system by the empirical potentials is important over the very wide temperatures induced during thermal spike and damage cascade simulations. This has motivated the development of a parameter set for gas-actinide and gas-oxygen interactions that is complementary for use with a recent many-body potential set. A comprehensive set of density functional theory (DFT) calculations were used to study Xe and Kr incorporation at a number of sites in CeO 2, ThO 2, UO 2 and PuO 2. These structures were used to fit a potential, which was used to generate molecular dynamics (MD) configurations incorporating Xe and Kr at 300 K, 1500 K, 3000 K and 5000 K. Subsequent matching to the forces predicted by DFT for these MD configurations was used to refine the potential set. This fitting approach ensured weighted fitting to configurations that are thermodynamically significant over a broad temperature range, while avoiding computationally expensive DFT-MD calculations. The resultant gas potentials were validated against DFT binding energies and are suitable for simulating combinations of Xe and Kr in solid solutions of CeO 2, ThO 2, UO 2 and PuO 2, providing a powerful tool for the atomistic simulation of conventional nuclear reactor fuel UO 2 as well as advanced MOX fuels.« less

Overview of Lockheed Martin cryocoolers

NASA Astrophysics Data System (ADS)

Nast, T.; Olson, J.; Champagne, P.; Evtimov, B.; Frank, D.; Roth, E.; Renna, T.

2006-02-01

Lockheed Martin's Advanced Technology Center (LM-ATC) in Palo Alto, California, has been active in space cryogenic developments for over 30 years. In prior years, work focused on stored cryogen systems for temperatures up to 125 K. As the mechanical cryocoolers matured and demonstrated reliable operation these stored cryogen systems gradually became replaced. LM-ATC is currently developing solid hydrogen systems for temperatures below 7 K [Naes L, Wu S, Cannon J. WISE solid hydrogen cryostat design overview. In: Proceedings of SPIE, cryogenic optical systems and instruments XI, vol. 5904, August, 2005], but these coolers will soon be replaced by mechanical cryocoolers. This paper will present a summary of cryocooler developments at LM-ATC and will describe the recent performance of multiple stage systems. A four-stage pulse tube cryocooler developed under contract to the Jet Propulsion Laboratory (JPL) has been recently developed and operated at 3.8 K [Olson JR, Moore M, Champagne P, Roth E, Evtimov B, Jensen J, et al. Development of a space-type-4-stage pulse tube cryocooler for very low temperatures, Adv Cryogen Engr, vol. 50, Amer Inst of Physics, New York, in press]. Coolers with one, two and three stages have also been widely developed [Nast TC et al. Miniature pulse tube cryocooler for space applications. Cryocoolers, vol. 11. New York: Plenum Press; 2000. p. 145-54; Olson J et al. Development of a 10 K pulse tube cryocooler for space applications. In: Ross R, editor. Cryocoolers, vol. 12. New York: Kluwer Academic/Plenum Publishers; 2003. p. 241-6; Nast TC et al. Lockheed Martin two-stage pulse tube cryocooler for GIFTS. Cryocoolers, vol. 13. New York: Kluwer Academic/Plenum Publishers; 2005; Frank D et al. Lockheed Martin RAMOS engineering model cryocooler. Cryocoolers, vol. 13. New York: Kluwer Academic/Plenum Publishers; 2005]. A staging approach is required to achieve very low temperatures, and also provides cooling at warmer temperatures, which is invariably beneficial in reducing heat loads to the lower temperature stages, or for cooling other system components. For example, our two-stage cooler [Nast TC et al. Lockheed Martin two-stage pulse tube cryocooler for GIFTS. Cryocoolers, vol. 13. New York: Kluwer Academic/Plenum Publishers; 2005; Frank D et al. Lockheed Martin RAMOS engineering model cryocooler. Cryocoolers, vol. 13. New York: Kluwer Academic/Plenum Publishers; 2005] is used to cool a low-temperature focal plane as well as a higher temperature optical sensor, using a single compressor and electronics at a substantial benefit in weight, reliability and cost.

Status report on the disposal of radioactive wastes

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

Culler, F.L. Jr.; McLain, S.

1957-06-25

A comprehensive survey of waste disposal techniques, requirements, costs, hazards, and long-range considerations is presented. The nature of high level wastes from reactors and chemical processes, in the form of fission product gases, waste solutions, solid wastes, and particulate solids in gas phase, is described. Growth predictions for nuclear reactor capacity and the associated fission product and transplutonic waste problem are made and discussed on the basis of present knowledge. Biological hazards from accumulated wastes and potential hazards from reactor accidents, ore and feed material processing, chemical reprocessing plants, and handling of fissionable and fertile material after irradiation and decontaminationmore » are surveyed. The waste transportation problem is considered from the standpoints of magnitude of the problem, present regulations, costs, and cooling periods. The possibilities for ultimate waste management and/or disposal are reviewed and discussed. The costs of disposal, evaporation, storage tanks, and drum-drying are considered.« less

Uranium plasma emission at gas-core reaction conditions

NASA Technical Reports Server (NTRS)

Williams, M. D.; Jalufka, N. W.; Hohl, F.; Lee, J. H.

1976-01-01

The results of uranium plasma emission produced by two methods are reported. For the first method a ruby laser was focused on the surface of a pure U-238 sample to create a plasma plume with a peak plasma density of about 10 to the 20th power/cu cm and a temperature of about 38,600 K. The absolute intensity of the emitted radiation, covering the range from 300 to 7000 A was measured. For the second method, the uranium plasma was produced in a 20 kilovolt, 25 kilojoule plasma-focus device. The 2.5 MeV neutrons from the D-D reaction in the plasma focus are moderated by polyethylene and induce fissions in the U-235. Spectra of both uranium plasmas were obtained over the range from 30 to 9000 A. Because of the low fission yield the energy input due to fissions is very small compared to the total energy in the plasma.

Environment control system

DOEpatents

Sammarone, Dino G.

1978-01-01

A system for controlling the environment of an enclosed area in nuclear reactor installations. The system permits the changing of the environment from nitrogen to air, or from air to nitrogen, without the release of any radioactivity or process gas to the outside atmosphere. In changing from a nitrogen to an air environment, oxygen is inserted into the enclosed area at the same rate which the nitrogen-oxygen gas mixture is removed from the enclosed area. The nitrogen-oxygen gas mixture removed from the enclosed area is mixed with hydrogen, the hydrogen recombining with the oxygen present in the gas to form water. The water is then removed from the system and, if it contains any radioactive products, can be utilized to form concrete, which can then be transferred to a licensed burial site. The process gas is purified further by stripping it of carbon dioxide and then distilling it to remove any xenon, krypton, and other fission or non-condensable gases. The pure nitrogen is stored as either a cryogenic liquid or a gas. In changing from an air to nitrogen environment, the gas is removed from the enclosed area, mixed with hydrogen to remove the oxygen present, dried, passed through adsorption beds to remove any fission gases, and reinserted into the enclosed area. Additionally, the nitrogen stored during the nitrogen to air change, is inserted into the enclosed area, the nitrogen from both sources being inserted into the enclosed area at the same rate as the removal of the gas from the containment area. As designed, the amount of nitrogen stored during the nitrogen to air change substantially equals that required to replace oxygen removed during an air to nitrogen change.

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.

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

Hu, Shenyang; Joshi, Vineet; Lavender, Curt A.

Experiments showed that recrystallization dramatically speeds up the gas bubble swelling kinetics in metallic UMo fuels. In this work a recrystallization model is developed to study the effect of microstructures and radiation conditions on recrystallization kinetics. The model integrates the rate theory of intra-granular gas bubble and interstitial loop evolution and a phase field model of recrystallization zone evolution. A fast passage method is employed to describe one dimensional diffusion of interstitials which have diffusivity several order magnitude larger than that of the fission gas Xe. With the model, the effect of grain sizes on recrystallization kinetics is simulated.

14. Elevation of boiler backhead showing (left to right at ...

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

14. Elevation of boiler backhead showing (left to right at top) steam pressure gauge, sight glass (indicates water level in boiler), manhole (for maintenance access to steam space), and try-cocks (used to determine water level if sight glass is inoperative). Below the firedoors lie air plenums which supply air from blower to firegrates; plenum door at lower left has been removed for photography. Each boiler was built by W. & A. Fletcher Co. to operate at 50 p.s.i. - Steamboat TICONDEROGA, Shelburne Museum Route 7, Shelburne, Chittenden County, VT

Piloted rich-catalytic lean-burn hybrid combustor

DOEpatents

Newburry, Donald Maurice

2002-01-01

A catalytic combustor assembly which includes, an air source, a fuel delivery means, a catalytic reactor assembly, a mixing chamber, and a means for igniting a fuel/air mixture. The catalytic reactor assembly is in fluid communication with the air source and fuel delivery means and has a fuel/air plenum which is coated with a catalytic material. The fuel/air plenum has cooling air conduits passing therethrough which have an upstream end. The upstream end of the cooling conduits is in fluid communication with the air source but not the fuel delivery means.

Measurement of fission product gases in the atmosphere

NASA Astrophysics Data System (ADS)

Schell, W. R.; Tobin, M. J.; Marsan, D. J.; Schell, C. W.; Vives-Batlle, J.; Yoon, S. R.

1997-01-01

The ability to quickly detect and assess the magnitude of releases of fission-produced radioactive material is of significant importance for ongoing operations of any conventional nuclear power plant or other activities with a potential for fission product release. In most instances, the control limits for the release of airborne radioactivity are low enough to preclude direct air sampling as a means of detection, especially for fission gases that decay by beta or electron emission. It is, therefore, customary to concentrate the major gaseous fission products (krypton, xenon and iodine) by cryogenic adsorption for subsequent separation and measurement. This study summarizes our initial efforts to develop an automated portable system for on-line separation and concentration with the potential for measuring environmental levels of radioactive gases, including 85Kr, 131,133,135Xe, 14C, 3H, 35S, 125,131I, etc., without using cryogenic fluids. Bench top and prototype models were constructed using the principle of heatless fractionation of the gases in a pressure swing system. This method removes the requirement for cryogenic fluids to concentrate gases and, with suitable electron and gamma ray detectors, provides for remote use under automatic computer control. Early results using 133Xe tracer show that kinetic chromatography, i.e., high pressure adsorption of xenon and low pressure desorption of air, using specific types of molecular sieves, permits the separation and quantification of xenon isotopes from large volume air samples. We are now developing the ability to measure the presence and amounts of fission-produced xenon isotopes that decay by internal conversion electrons and beta radiation with short half-lives, namely 131mXe, 11.8 d, 133mXe, 2.2 d, 133Xe, 5.2 d and 135Xe, 9.1 h. The ratio of the isotopic concentrations measured can be used to determine unequivocally the amount of fission gas and time of release of an air parcel many kilometers downwind from a nuclear activity where the fission products were discharged.

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

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

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

The simulation of air recirculation and fire/explosion phenomena within a semiconductor factory.

PubMed

I, Yet-Pole; Chiu, Yi-Long; Wu, Shi-Jen

2009-04-30

The semiconductor industry is the collection of capital-intensive firms that employ a variety of hazardous chemicals and engage in the design and fabrication of semiconductor devices. Owing to its processing characteristics, the fully confined structure of the fabrication area (fab) and the vertical airflow ventilation design restrict the applications of traditional consequence analysis techniques that are commonly used in other industries. The adverse situation also limits the advancement of a fire/explosion prevention design for the industry. In this research, a realistic model of a semiconductor factory with a fab, sub-fabrication area, supply air plenum, and return air plenum structures was constructed and the computational fluid dynamics algorithm was employed to simulate the possible fire/explosion range and its severity. The semiconductor factory has fan module units with high efficiency particulate air filters that can keep the airflow uniform within the cleanroom. This condition was modeled by 25 fans, three layers of porous ceiling, and one layer of porous floor. The obtained results predicted very well the real airflow pattern in the semiconductor factory. Different released gases, leak locations, and leak rates were applied to investigate their influence on the hazard range and severity. Common mitigation measures such as a water spray system and a pressure relief panel were also provided to study their potential effectiveness to relieve thermal radiation and overpressure hazards within a fab. The semiconductor industry can use this simulation procedure as a reference on how to implement a consequence analysis for a flammable gas release accident within an air recirculation cleanroom.

Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

NASA Astrophysics Data System (ADS)

Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

2013-09-01

Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

Parents of two-phase flow and theory of "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper gives a brief overview of types of two-phase flow. Subsequently, it deals with their mutual division and problems with accuracy boundaries among particular types. It also shows the case of water flow through a pipe with external heating and the gradual origination of all kinds of flow. We have met it in solution of safety condition of various stages in pressurized and boiling water reactors. In the MSR there is a problem in the solution of gas-lift using helium as a gas and its secondary usage for clearing of the fuel mixture from gaseous fission products. Theory of gas-lift is described.

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

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

Barani, T.; Bruschi, E.; Pizzocri, D.

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which ismore » applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.« less

Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

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

Wichner, R.P.; Hodge, S.A.; Weber, C.F.

1984-08-01

This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

Neutron induced fission cross section measurements of 240Pu and 242Pu

NASA Astrophysics Data System (ADS)

Belloni, F.; Eykens, R.; Heyse, J.; Matei, C.; Moens, A.; Nolte, R.; Plompen, A. J. M.; Richter, S.; Sibbens, G.; Vanleeuw, D.; Wynants, R.

2017-09-01

Accurate neutron induced fission cross section of 240Pu and 242Pu are required in view of making nuclear technology safer and more efficient to meet the upcoming needs for the future generation of nuclear power plants (GEN-IV). The probability for a neutron to induce such reactions figures in the NEA Nuclear Data High Priority Request List [1]. A measurement campaign to determine neutron induced fission cross sections of 240Pu and 242Pu at 2.51 MeV and 14.83 MeV has been carried out at the 3.7 MV Van De Graaff linear accelerator at Physikalisch-Technische Bundesanstalt (PTB) in Braunschweig. Two identical Frisch Grid fission chambers, housing back to back a 238U and a APu target (A = 240 or A = 242), were employed to detect the total fission yield. The targets were molecular plated on 0.25 mm aluminium foils kept at ground potential and the employed gas was P10. The neutron fluence was measured with the proton recoil telescope (T1), which is the German primary standard for neutron fluence measurements. The two measurements were related using a De Pangher long counter and the charge as monitors. The experimental results have an average uncertainty of 3-4% at 2.51 MeV and for 6-8% at 14.81 MeV and have been compared to the data available in literature.

Postirradiation analysis of the latest high uranium density miniplate test: RERTR 8.

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

Hofman, G. L.; Kim, Y. S.; Rest, J.

2008-01-01

Results of destructive examination of fuel miniplates irradiated in the RERTR-8 test are discussed. Metallographic features of dispersion fuel containing fuel particles of U-7wt%Mo with 1wt% Ti or 2wt% Zr are analyzed. It is hypothesized that Zr, either as alloy addition or fission product, may have a destabilizing effect on fission gas behavior. The purpose of miniplate test RERTR-8 was to obtain irradiation performance data on monolithic fuel plates fabricated by friction bonding (FB) and isostatic hot pressing (HIP), as well as dispersion fuel plates that contain U-7Mo fuel particles alloyed with small amounts of Zr or Ti (see Fig.more » 1). The results of the monolithic plates destructively examined to date were presented at the 2007 RERTR meeting in Prague. This paper presents the first results on the dispersion plates with Ti and Zr additions to U-7Mo. The effect of Ti and Zr additions to U-7wt%Mo on the extent of fuel-aluminum interdiffusion, although measureable, is small in absolute terms because of the overwhelming effect of the 5% Si addition to the Al matrix. Ti additions to the U-7wt%Mo have no discernable effect on swelling behavior of the fuel. However, there are indications that the addition of Zr may have a destabilizing effect on fission gas behavior at high burnup.« less

Weld Development for Aluminum Fission Chamber

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

Cross, Carl Edward; Martinez, Jesse Norris

2017-05-16

The Sigma welding team was approached to help fabricate a small fission chamber (roughly ½ inch dia. x ½ inch tall cylinder). These chambers are used as radiation sensors that contain small traces of radionuclides (Cf 252, U 235, and U 238) that serve to ionize gas atoms in addition to external radiation. When a voltage is applied within the chamber, the resulting ion flow can be calibrated and monitored. Aluminum has the advantage of not forming radioactive compounds when exposed to high external radiation (except from minor Na alloy content). Since aluminum has not been used before in thismore » application, this presented an unexplored challenge.« less

Fission gas in thoria

NASA Astrophysics Data System (ADS)

Kuganathan, Navaratnarajah; Ghosh, Partha S.; Galvin, Conor O. T.; Arya, Ashok K.; Dutta, Bijon K.; Dey, Gautam K.; Grimes, Robin W.

2017-03-01

The fission gases Xe and Kr, formed during normal reactor operation, are known to degrade fuel performance, particularly at high burn-up. Using first-principles density functional theory together with a dispersion correction (DFT + D), in ThO2 we calculate the energetics of neutral and charged point defects, the di-vacancy (DV), different neutral tri-vacancies (NTV), the charged tetravacancy (CTV) defect cluster geometries and their interaction with Xe and Kr. The most favourable incorporation point defect site for Xe or Kr in defective ThO2 is the fully charged thorium vacancy. The lowest energy NTV in larger supercells of ThO2 is NTV3, however, a single Xe atom is most stable when accommodated within a NTV1. The di-vacancy (DV) is a significantly less favoured incorporation site than the NTV1 but the CTV offers about the same incorporation energy. Incorporation of a second gas atom in a NTV is a high energy process and more unfavourable than accommodation within an existing Th vacancy. The bi-NTV (BNTV) cluster geometry studied will accommodate one or two gas atoms with low incorporation energies but the addition of a third gas atom incurs a high energy penalty. The tri-NTV cluster (TNTV) forms a larger space which accommodates three gas atoms but again there is a penalty to accommodate a fourth gas atom. By considering the energy to form the defect sites, solution energies were generated showing that in ThO2-x the most favourable solution equilibrium site is the NTV1 while in ThO2 it is the DV.

Phase 1 Space Fission Propulsion Energy Source Design

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems with a specific mass at or below 50 kg/kWjet could enhance or enable numerous robotic outer solar system missions of interest. At the required specific mass, it is possible to develop safe, affordable systems that meet mission requirements. To help select the system design to pursue, eight evaluation criteria were identified: system integration, safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of four potential concepts was performed: a Testable, Passive, Redundant Reactor (TPRR), a Testable Multi-Cell In-Core Thermionic Reactor (TMCT), a Direct Gas Cooled Reactor (DGCR), and a Pumped Liquid Metal Reactor.(PLMR). Development of any of the four systems appears feasible. However, for power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the TPRR has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the TPRR approach. Successful development and utilization of a "Phase I" fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system.

Energy release, beam attenuation radiation damage, gas production and accumulation of long-lived activity in Pb, Pb-Bi and Hg targets

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

Shubin, Yu.N.

1996-06-01

The calculation and analysis of the nuclei concentrations and long-lived residual radioactivity accumulated in Pb, Pb-Bi and Hg targets irradiated by 800 MeV, 30 mA proton beam have been performed. The dominating components to the total radioactivity of radionuclides resulting from fission and spallation reactions and radiative capture by both target nuclei and accumulated radioactive nuclei for various irradiation and cooling times were analyzed. The estimations of spectral component contributions of neutron and proton fluxes to the accumulated activity were carried out. The contributions of fission products to the targets activity and partial activities of main long-lived fission products tomore » the targets activity and partial activities of main long-lived fission products were evaluated. The accumulation of Po isotopes due to reactions induced by secondary alpha-particles were found to be important for the Pb target as compared with two-step radiative capture. The production of Tritium in the targets and its contribution to the total targets activity was considered in detail. It is found that total activities of both targets are close to one another.« less

A two-dimensional adaptive-wall test section with ventilated walls in the Ames 2- by 2-foot transonic wind tunnel

NASA Technical Reports Server (NTRS)

Schairer, Edward T.; Lee, George; Mcdevitt, T. Kevin

1989-01-01

The first tests conducted in the adaptive-wall test section of the Ames Research Center's 2- by 2-Foot Transonic Wind Tunnel are described. A procedure was demonstrated for reducing wall interference in transonic flow past a two-dimensional airfoil by actively controlling flow through the slotted walls of the test section. Flow through the walls was controlled by adjusting pressures in compartments of plenums above and below the test section. Wall interference was assessed by measuring (with a laser velocimeter) velocity distributions along a contour surrounding the model, and then checking those measurements for their compatibility with free-air far-field boundary conditions. Plenum pressures for minimum wall interference were determined from empirical influence coefficients. An NACA 0012 airfoil was tested at angles of attach of 0 and 2, and at Mach numbers between 0.70 and 0.85. In all cases the wall-setting procedure greatly reduced wall interference. Wall interference, however, was never completely eliminated, primarily because the effect of plenum pressure changes on the velocities along the contour could not be accurately predicted.

A Transonic and Surpersonic Investigation of Jet Exhaust Plume Effects on the Afterbody and Base Pressures of a Body of Revolution

NASA Technical Reports Server (NTRS)

Andrews, C. D.; Cooper, C. E., Jr.

1974-01-01

An experimental aerodynamic investigation was conducted to provide data for studies to determine the criteria for simulating rocket engine plume induced aerodynamic effects in the wind tunnel using a simulated gaseous plume. Model surface and base pressure data were obtained in the presence of both a simulated and a prototype gaseous plume for a matrix of plume properties to enable investigators to determine the parameters that correlate the simulated and prototype plume-induced data. The test program was conducted in the Marshall Space Flight Center's 14 x 14-inch trisonic wind tunnel using two models, the first being a strut mounted cone-ogive-cylinder model with a fineness ratio of 9. Model exterior pressures, model plenum chamber and nozzle performance data were obtained at Mach numbers of 0.9, 1.2, 1.46, and 3.48. The exhaust plume was generated by using air as the simulant gas, or Freon-14 (CF4) as the prototype gas, over a chamber pressure range from 0 to 2,000 psia and a total temperature range from 50 to 600 F.

Exit chimney joint and method of forming the joint for closed circuit steam cooled gas turbine nozzles

DOEpatents

Burdgick, Steven Sebastian; Burns, James Lee

2002-01-01

A nozzle segment for a gas turbine includes inner and outer band portions and a vane extending between the band portions. The inner and outer band portions are each divided into first and second plenums separated by an impingement plate. Cooling steam is supplied to the first cavity for flow through the apertures to cool the outer nozzle wall. The steam flows through a leading edge cavity in the vane into the first cavity of the inner band portion for flow through apertures of the impingement plate to cool the inner nozzle wall. Spent cooling steam flows through a plurality of cavities in the vane, exiting through an exit chimney in the outer band. The exit chimney is secured at its inner end directly to the nozzle vane wall surrounding the exit cavities, to the margin of the impingement plate at a location intermediate the ends of the exit chimney and to margins of an opening through the cover whereby each joint is externally accessible for joint formation and for subsequent inspection.

The Eastern Gas Shales Project (EGSP) Data System: A case study in data base design, development, and application

USGS Publications Warehouse

Dyman, T.S.; Wilcox, L.A.

1983-01-01

The U.S. Geological Survey and Petroleum Information Corporation in Denver, Colorado, developed the Eastern Gas Shale Project (EGSP)Data System for the U.S. Department of Energy, Morgantown, West Virginia. Geological, geochemical, geophysical, and engineering data from Devonian shale samples from more than 5800 wells and outcrops in the Appalachian basin were edited and converted to a Petroleum Information Corporation data base. Well and sample data may be retrieved from this data system to produce (1)production-test summaries by formation and well location; (2)contoured isopach, structure, and trendsurface maps of Devonian shale units; (3)sample summary reports for samples by location, well, contractor, and sample number; (4)cross sections displaying digitized log traces, geochemical, and lithologic data by depth for wells; and (5)frequency distributions and bivariate plots. Although part of the EGSP Data System is proprietary, and distribution of complete well histories is prohibited by contract, maps and aggregated well-data listings are being made available to the public through published reports. ?? 1983 Plenum Publishing Corporation.

WELDED JACKETED URANIUM BODY

DOEpatents

Gurinsky, D.H.

1958-08-26

A fuel element is presented for a neutronic reactor and is comprised of a uranium body, a non-fissionable jacket surrounding sald body, thu jacket including a portion sealed by a weld, and an inclusion in said sealed jacket at said weld of a fiux having a low neutron capture cross-section. The flux is provided by combining chlorine gas and hydrogen in the intense heat of-the arc, in a "Heliarc" welding muthod, to form dry hydrochloric acid gas.

Thermodynamic and fluid mechanic analysis of rapid pressurization in a dead-end tube

NASA Technical Reports Server (NTRS)

Leslie, Ian H.

1989-01-01

Three models have been applied to very rapid compression of oxygen in a dead-ended tube. Pressures as high as 41 MPa (6000 psi) leading to peak temperatures of 1400 K are predicted. These temperatures are well in excess of the autoignition temperature (750 K) of teflon, a frequently used material for lining hoses employed in oxygen service. These findings are in accord with experiments that have resulted in ignition and combustion of the teflon, leading to the combustion of the stainless steel braiding and catastrophic failure. The system analyzed was representative of a capped off-high-pressure oxygen line, which could be part of a larger system. Pressurization of the larger system would lead to compression in the dead-end line, and possible ignition of the teflon liner. The model consists of a large plenum containing oxygen at the desired pressure (500 to 6000 psi). The plenum is connected via a fast acting valve to a stainless steel tube 2 cm inside diameter. Opening times are on the order of 15 ms. Downstream of the valve is an orifice sized to increase filling times to around 100 ms. The total length from the valve to the dead-end is 150 cm. The distance from the valve to the orifice is 95 cm. The models describe the fluid mechanics and thermodynamics of the flow, and do not include any combustion phenomena. A purely thermodynamic model assumes filling to be complete upstream of the orifice before any gas passes through the orifice. This simplification is reasonable based on experiment and computer modeling. Results show that peak temperatures as high as 4800 K can result from recompression of the gas after expanding through the orifice. An approximate transient model without an orifice was developed assuming an isentropic compression process. An analytical solution was obtained. Results indicated that fill times can be considerably shorter than valve opening times. The third model was a finite difference, 1-D transient compressible flow model. Results from the code show the recompression effect but predict much lower peak temperatures than the thermodynamic model.

Thermodynamic and fluid mechanic analysis of rapid pressurization in a dead-end tube

NASA Astrophysics Data System (ADS)

Leslie, Ian H.

1989-12-01

Three models have been applied to very rapid compression of oxygen in a dead-ended tube. Pressures as high as 41 MPa (6000 psi) leading to peak temperatures of 1400 K are predicted. These temperatures are well in excess of the autoignition temperature (750 K) of teflon, a frequently used material for lining hoses employed in oxygen service. These findings are in accord with experiments that have resulted in ignition and combustion of the teflon, leading to the combustion of the stainless steel braiding and catastrophic failure. The system analyzed was representative of a capped off-high-pressure oxygen line, which could be part of a larger system. Pressurization of the larger system would lead to compression in the dead-end line, and possible ignition of the teflon liner. The model consists of a large plenum containing oxygen at the desired pressure (500 to 6000 psi). The plenum is connected via a fast acting valve to a stainless steel tube 2 cm inside diameter. Opening times are on the order of 15 ms. Downstream of the valve is an orifice sized to increase filling times to around 100 ms. The total length from the valve to the dead-end is 150 cm. The distance from the valve to the orifice is 95 cm. The models describe the fluid mechanics and thermodynamics of the flow, and do not include any combustion phenomena. A purely thermodynamic model assumes filling to be complete upstream of the orifice before any gas passes through the orifice. This simplification is reasonable based on experiment and computer modeling. Results show that peak temperatures as high as 4800 K can result from recompression of the gas after expanding through the orifice. An approximate transient model without an orifice was developed assuming an isentropic compression process. An analytical solution was obtained.

Fuel nozzle tube retention

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

Cihlar, David William; Melton, Patrick Benedict

A system for retaining a fuel nozzle premix tube includes a retention plate and a premix tube which extends downstream from an outlet of a premix passage defined along an aft side of a fuel plenum body. The premix tube includes an inlet end and a spring support feature which is disposed proximate to the inlet end. The premix tube extends through the retention plate. The spring retention feature is disposed between an aft side of the fuel plenum and the retention plate. The system further includes a spring which extends between the spring retention feature and the retention plate.

Compressor bleed cooling fluid feed system

DOEpatents

Donahoo, Eric E; Ross, Christopher W

2014-11-25

A compressor bleed cooling fluid feed system for a turbine engine for directing cooling fluids from a compressor to a turbine airfoil cooling system to supply cooling fluids to one or more airfoils of a rotor assembly is disclosed. The compressor bleed cooling fluid feed system may enable cooling fluids to be exhausted from a compressor exhaust plenum through a downstream compressor bleed collection chamber and into the turbine airfoil cooling system. As such, the suction created in the compressor exhaust plenum mitigates boundary layer growth along the inner surface while providing flow of cooling fluids to the turbine airfoils.

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

NASA Technical Reports Server (NTRS)

Springborn, R. H.

1971-01-01

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

High flux reactor

DOEpatents

Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

1988-01-01

A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

Free-carrier-induced soliton fission unveiled by in situ measurements in nanophotonic waveguides

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

Husko, Chad; Wulf, Matthias; Lefrancois, Simon

Solitons are localized waves formed by a balance of focusing and defocusing effects. These nonlinear waves exist in diverse forms of matter yet exhibit similar properties including stability, periodic recurrence and particle-like trajectories. One important property is soliton fission, a process by which an energetic higher-order soliton breaks apart due to dispersive or nonlinear perturbations. Here we demonstrate through both experiment and theory that nonlinear photocarrier generation can induce soliton fission. Using near-field measurements, we directly observe the nonlinear spatial and temporal evolution of optical pulses in situ in a nanophotonic semiconductor waveguide. We develop an analytic formalism describing themore » free-carrier dispersion (FCD) perturbation and show the experiment exceeds the minimum threshold by an order of magnitude. We confirm these observations with a numerical nonlinear Schrodinger equation model. Finally, these results provide a fundamental explanation and physical scaling of optical pulse evolution in free-carrier media and could enable improved supercontinuum sources in gas based and integrated semiconductor waveguides.« less

Free-carrier-induced soliton fission unveiled by in situ measurements in nanophotonic waveguides

DOE PAGES

Husko, Chad; Wulf, Matthias; Lefrancois, Simon; ...

2016-04-15

Solitons are localized waves formed by a balance of focusing and defocusing effects. These nonlinear waves exist in diverse forms of matter yet exhibit similar properties including stability, periodic recurrence and particle-like trajectories. One important property is soliton fission, a process by which an energetic higher-order soliton breaks apart due to dispersive or nonlinear perturbations. Here we demonstrate through both experiment and theory that nonlinear photocarrier generation can induce soliton fission. Using near-field measurements, we directly observe the nonlinear spatial and temporal evolution of optical pulses in situ in a nanophotonic semiconductor waveguide. We develop an analytic formalism describing themore » free-carrier dispersion (FCD) perturbation and show the experiment exceeds the minimum threshold by an order of magnitude. We confirm these observations with a numerical nonlinear Schrodinger equation model. Finally, these results provide a fundamental explanation and physical scaling of optical pulse evolution in free-carrier media and could enable improved supercontinuum sources in gas based and integrated semiconductor waveguides.« less

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

Ronchi, C.; Sari, C.

Lenticular pore migration rates in oxide nuclear fuels were measured in out-of-pile heating experiments. It is deduced that those pores which are in part responsible for the formation of columnar grains, are only produced in the absence of relevant amourts of filling gas. Specimens containing important concentrations of He, produced by Pu alpha decay, show columnar grain restructuring by grain boundary migration. Some consequences are drawn concerning the possible role played by lenticular pores in the mechanisms of fission gas release from nuclear fuels. (5 figures) (auth)

Connecting the "Hot Fusion Island" to the Nuclear Mainland: Search for 283,284,285Fl Decay Chains

NASA Astrophysics Data System (ADS)

Rykaczewski, K. P.; Utyonkov, V. K.; Brewer, N. T.; Grzywacz, R. K.; Miernik, K.; Roberto, J. B.; Oganessian, Yu. Ts.; Polyakov, A. N.; Tsyganov, Yu. S.; Voinov, A. A.; Abdullin, F. Sh.; Dmitriev, S. N.; Itkis, M. G.; Sabelnikov, A. V.; Sagaidak, R. N.; Shirokovsky, I. V.; Shumeiko, M. V.; Subbotin, V. G.; Sukhov, A. M.; Vostokin, G. K.; Hamilton, J. H.; Henderson, R. A.; Stoyer, M. A.

The program of studies on superheavy nuclei to identify new isotopes anchoring the decay chains from the Hot Fusion Island to the Nuclear Mainland has been started at the Dubna Gas Filled Recoil Separator (DGFRS, JINR Dubna) in collaboration between Russia, US and Poland. These studies are performed with new detection and digital data acquisition system developed at ORNL (Oak Ridge) and UT (Knoxville). The evidence for fast fission of the new isotope 284Fl is presented. The low cross section for the 3n channel of 239Pu + 48Ca reaction is attributed to lower than expected fission barriers in 287-284Fl.

Radiant vessel auxiliary cooling system

DOEpatents

Germer, John H.

1987-01-01

In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

Segmented annular combustor

DOEpatents

Reider, Samuel B.

1979-01-01

An industrial gas turbine engine includes an inclined annular combustor made up of a plurality of support segments each including inner and outer walls of trapezoidally configured planar configuration extents and including side flanges thereon interconnected by means of air cooled connector bolt assemblies to form a continuous annular combustion chamber therebetween and wherein an air fuel mixing chamber is formed at one end of the support segments including means for directing and mixing fuel within a plenum and a perforated header plate for directing streams of air and fuel mixture into the combustion chamber; each of the outer and inner walls of each of the support segments having a ribbed lattice with tracks slidably supporting porous laminated replaceable panels and including pores therein for distributing combustion air into the combustion chamber while cooling the inner surface of each of the panels by transpiration cooling thereof.

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

Advanced Space Fission Propulsion Systems

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Borowski, Stanley K.

2010-01-01

Fission has been considered for in-space propulsion since the 1940s. Nuclear Thermal Propulsion (NTP) systems underwent extensive development from 1955-1973, completing 20 full power ground tests and achieving specific impulses nearly twice that of the best chemical propulsion systems. Space fission power systems (which may eventually enable Nuclear Electric Propulsion) have been flown in space by both the United States and the Former Soviet Union. Fission is the most developed and understood of the nuclear propulsion options (e.g. fission, fusion, antimatter, etc.), and fission has enjoyed tremendous terrestrial success for nearly 7 decades. Current space nuclear research and technology efforts are focused on devising and developing first generation systems that are safe, reliable and affordable. For propulsion, the focus is on nuclear thermal rockets that build on technologies and systems developed and tested under the Rover/NERVA and related programs from the Apollo era. NTP Affordability is achieved through use of previously developed fuels and materials, modern analytical techniques and test strategies, and development of a small engine for ground and flight technology demonstration. Initial NTP systems will be capable of achieving an Isp of 900 s at a relatively high thrust-to-weight ratio. The development and use of first generation space fission power and propulsion systems will provide new, game changing capabilities for NASA. In addition, development and use of these systems will provide the foundation for developing extremely advanced power and propulsion systems capable of routinely and affordably accessing any point in the solar system. The energy density of fissile fuel (8 x 10(exp 13) Joules/kg) is more than adequate for enabling extensive exploration and utilization of the solar system. For space fission propulsion systems, the key is converting the virtually unlimited energy of fission into thrust at the desired specific impulse and thrust-to-weight ratio. This presentation will discuss potential space fission propulsion options ranging from first generation systems to highly advanced systems. Ongoing research that shows promise for enabling second generation NTP systems with Isp greater than 1000 s will be discussed, as will the potential for liquid, gas, or plasma core systems. Space fission propulsion systems could also be used in conjunction with simple (water-based) propellant depots to enable routine, affordable missions to various destinations (e.g. moon, Mars, asteroids) once in-space infrastructure is sufficiently developed. As fuel and material technologies advance, very high performance Nuclear Electric Propulsion (NEP) systems may also become viable. These systems could enable sophisticated science missions, highly efficient cargo delivery, and human missions to numerous destinations. Commonalities between NTP, fission power systems, and NEP will be discussed.

Investigations on the self-excited oscillations in a kerosene spray flame

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

de la Cruz Garcia, M.; Mastorakos, E.; Dowling, A.P.

2009-02-15

A laboratory scale gas turbine type burner at atmospheric pressure and with air preheat was operated with aviation kerosene Jet-A1 injected from a pressure atomiser. Self-excited oscillations were observed and analysed to understand better the relationship between the spray and thermo-acoustic oscillations. The fluctuations of CH{sup *} chemiluminescence measured simultaneously with the pressure were used to determine the flame transfer function. The Mie scattering technique was used to record spray fluctuations in reacting conditions with a high speed camera. Integrating the Mie intensity over the imaged region gave a temporal signal acquired simultaneously with pressure fluctuations and the transfer functionmore » between the light scattered from the spray and the velocity fluctuations in the plenum was evaluated. Phase Doppler anemometry was used for axial velocity and drop size measurements at different positions downstream the injection plane and for various operating conditions. Pressure spectra showed peaks at a frequency that changed with air mass flow rate. The peak for low air mass flow rate operation was at 220 Hz and was associated with a resonance of the supply plenum. At the same global equivalence ratio but at high air mass flow rates, the pressure spectrum peak was at 323 Hz, a combustion chamber resonant frequency. At low air flow rates, the spray fluctuation motion was pronounced and followed the frequency of the pressure oscillation. At high air flow rates, more effective evaporation resulted in a complete disappearance of droplets at an axial distance of about 1/3 burner diameters from the injection plane, leading to a different flame transfer function and frequency of the self-excited oscillation. The results highlight the sensitivity of the self-excited oscillation to the degree of mixing achieved before the main recirculation zone. (author)« less

Aircraft anti-insect system

NASA Technical Reports Server (NTRS)

Spiro, Clifford Lawrence (Inventor); Fric, Thomas Frank (Inventor); Leon, Ross Michael (Inventor)

1997-01-01

Insect debris is removed from or prevented from adhering to insect impingement areas of an aircraft, particularly on an inlet cowl of an engine, by heating the area to 180.degree.-500.degree. C. An apparatus comprising a means to bring hot air from the aircraft engine to a plenum contiguous to the insect impingement area provides for the heating of the insect impingement areas to the required temperatures. The plenum can include at least one tube with a plurality of holes contained in a cavity within the inlet cowl. It can also include an envelope with a plurality of holes on its surface contained in a cavity within the inlet cowl.

Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

NASA Astrophysics Data System (ADS)

Sabotinov, Luben; Chevrier, Patrick

The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

Capabilities and Testing of the Fission Surface Power Primary Test Circuit (FSP-PTC)

NASA Technical Reports Server (NTRS)

Garber, Anne E.

2007-01-01

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK), which was used in the SNAP-10A fission reactor, was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around a 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. NaK flow rates of greater than 1 kg/sec may be achieved, depending upon the power applied to the EM pump. The heat exchanger provides for the removal of thermal energy from the circuit, simulating the presence of an energy conversion system. The presence of the test section increases the versatility of the circuit. A second liquid metal pump, an energy conversion system, and highly instrumented thermal simulators are all being considered for inclusion within the test section. This paper summarizes the capabilities and ongoing testing of the Fission Surface Power Primary Test Circuit (FSP-PTC).

Search for Cm-248 in the early solar system

NASA Technical Reports Server (NTRS)

Lavielle, B.; Marti, K.; Pellas, P.; Perron, C.

1992-01-01

Possible evidence for the presence of Cm-248 in the early solar system was reported from fission gas studies (Rao and Gopalan, 1973) and recently from studies of very high nuclear track densities (not less than 5 x 10 exp 8/sq cm) in the merrillite of the H4 chondrite Forest Vale (F.V.) (Pellas et al., 1987). We report here an analysis of the isotopic abundances of xenon in F.V. phosphates and results of track studies in phosphate/pyroxene contacts. The fission xenon isotopic signature clearly identifies Pu-244 as the extinct progenitor. We calculate an upper limit Cm-248/Pu-244 to be less than 0.0015 at the beginning of Xe retention in F.V. phosphates. This corresponds to an upper limit of the ratio Cm-248/U-235 of not greater than 5 x 10 exp -5 further constraining the evidence for any late addition of freshly synthesized actinide elements just prior to solar system formation. The fission track density observed after annealing the phosphates at 290C (1 hr, which essentially erases spallation recoil tracks) is also in agreement with the Pu-244 abundance inferred from fission Xe. The spallation recoil tracks produced during the 76 Ma cosmic-ray exposure account for the very high track density in merrillites.

Measurements of charge distributions of the fragments in the low energy fission reaction

NASA Astrophysics Data System (ADS)

Wang, Taofeng; Han, Hongyin; Meng, Qinghua; Wang, Liming; Zhu, Liping; Xia, Haihong

2013-01-01

The measurement for charge distributions of fragments in spontaneous fission 252Cf has been performed by using a unique style of detector setup consisting of a typical grid ionization chamber and a ΔΕ-Ε particle telescope, in which a thin grid ionization chamber served as the ΔΕ-section and the E-section was an Au-Si surface barrier detector. The typical physical quantities of fragments, such as mass number and kinetic energies as well as the deposition in the gas ΔΕ detector and E detector were derived from the coincident measurement data. The charge distributions of the light fragments for the fixed mass number A2* and total kinetic energy (TKE) were obtained by the least-squares fits for the response functions of the ΔΕ detector with multi-Gaussian functions representing the different elements. The results of the charge distributions for some typical fragments are shown in this article which indicates that this detection setup has the charge distribution capability of Ζ:ΔΖ>40:1. The experimental method developed in this work for determining the charge distributions of fragments is expected to be employed in the neutron induced fissions of 232Th and 238U or other low energy fission reactions.

LaRC results on nuclear pumped noble gas lasers

NASA Technical Reports Server (NTRS)

Deyoung, R. J.

1979-01-01

The recent experiment and theoretical results obtained for noble gas nuclear laser systems are presented. It is shown that the noble gas lasers are among the easiest systems to pump by nuclear excitation and as a result, all of the noble gases except He have lased under nuclear excitation. The noble gas systems are not ideal for high-power applications but they do give valuable insight into the operation and pumping mechanisms associated with nuclear lasers. At present, the Ar-Xe system is the best noble gas candidate for (U-235)F6 pumping. It appears that the quenching of Ar-Xe lasing is a result of the fluorine and not the uranium or fission fragments themselves. Thus, to achieve lasing with UF6, a fluorine compatible system must be found.

Multi-Megawatt Gas Turbine Power Systems for Lunar Colonies

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.

2006-01-01

A concept for development of second generation 10 MWe prototype lunar power plant utilizing a gas cooled fission reactor supplying heated helium working fluid to two parallel 5 MWe closed cycle gas turbines is presented. Such a power system is expected to supply the energy needs for an initial lunar colony with a crew of up to 50 persons engaged in mining and manufacturing activities. System performance and mass details were generated by an author developed code (BRMAPS). The proposed pilot power plant can be a model for future plants of the same capacity that could be tied to an evolutionary lunar power grid.

Fission Surface Power Technology Development Update

NASA Technical Reports Server (NTRS)

Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

2011-01-01

Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power conversion unit with electrical controls, and a heat rejection system with a multi-panel radiator assembly. Testing is planned at the Glenn Research Center Vacuum Facility 6 starting in 2012, with vacuum and liquid-nitrogen cold walls to provide simulation of operationally relevant environments. A nominal two-year test campaign is planned including a Phase 1 reactor simulator and power conversion test followed by a Phase 2 integrated system test with radiator panel heat rejection. The testing is expected to demonstrate the readiness and availability of fission surface power as a viable power system option for NASA's exploration needs. In addition to surface power, technology development work within this project is also directly applicable to in-space fission power and propulsion systems.

Energy 83. Revised and Expanded.

ERIC Educational Resources Information Center

Lord, John, Ed.

Energy 80 is an energy education program for middle/junior high school students. This document is a booklet of energy topics designed for student use in the program. Topics considered in this booklet include: forms of energy; energy rules; solar energy; food energy; fossil fuels; coal; oil and gas production and consumption; nuclear fission;…

Energy 80 for the 1981-82 School Year. [Student Handbook].

ERIC Educational Resources Information Center

Enterprise for Education, Santa Monica, CA.

Energy 80 is a booklet of energy topics for junior/high/middle school students. The topics are presented in 16 short sections (spreads). Topics include: energy forms; energy rules; solar energy; food energy; origin of fossil fuels; coal; oil and gas production and consumption; nuclear fission; renewable energy sources; history of United States…

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

Evaluation of design parameters for TRISO-coated fuel particles to establish manufacturing critical limits using PARFUME

DOE PAGES

Skerjanc, William F.; Maki, John T.; Collin, Blaise P.; ...

2015-12-02

The success of modular high temperature gas-cooled reactors is highly dependent on the performance of the tristructural-isotopic (TRISO) coated fuel particle and the quality to which it can be manufactured. During irradiation, TRISO-coated fuel particles act as a pressure vessel to contain fission gas and mitigate the diffusion of fission products to the coolant boundary. The fuel specifications place limits on key attributes to minimize fuel particle failure under irradiation and postulated accident conditions. PARFUME (an integrated mechanistic coated particle fuel performance code developed at the Idaho National Laboratory) was used to calculate fuel particle failure probabilities. By systematically varyingmore » key TRISO-coated particle attributes, failure probability functions were developed to understand how each attribute contributes to fuel particle failure. Critical manufacturing limits were calculated for the key attributes of a low enriched TRISO-coated nuclear fuel particle with a kernel diameter of 425 μm. As a result, these critical manufacturing limits identify ranges beyond where an increase in fuel particle failure probability is expected to occur.« less

Electron-beam-driven RI separator for SCRIT (ERIS) at RIKEN RI beam factory

NASA Astrophysics Data System (ADS)

Ohnishi, T.; Ichikawa, S.; Koizumi, K.; Kurita, K.; Miyashita, Y.; Ogawara, R.; Tamaki, S.; Togasaki, M.; Wakasugi, M.

2013-12-01

We constructed a radioactive isotope (RI) separator named ERIS (electron-beam-driven RI separator for SCRIT) for the SCRIT (Self-Confinement RI Target) electron scattering facility at RIKEN RI Beam Factory (RIBF). In ERIS, production rate of fission products in the photofission of uranium is estimated to be 2.2 ×1011 fissions/s with 30 g of uranium and a 1-kW electron beam. During the commissioning of ERIS, the mass resolution and overall efficiency, including ionization, extraction, and transmission, were found to be 1660 and 21%, respectively, using natural xenon gas. The preparation of uranium carbide (UC2) RI production targets is described from which a 132Sn beam was successfully separated in our first attempt at RI production.

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

NASA Astrophysics Data System (ADS)

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; Hunn, John D.; Baldwin, Charles A.; Montgomery, Fred C.; Terry, Jeff

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

DOE PAGES

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; ...

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

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

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

American Security and the International Energy Situation. Volume 2. World Energy and the Security of Supply

DTIC Science & Technology

1975-04-15

flue gas desulfurization technology seems to oe progressing so that by the late 1970s utilities may be able to burn high-sultur coal directly with...CObHqat ion•.V Conferva 1i on 0’ I , gas . and shale Coa I Lir.’I ronmcntal control Nuclear fission Nuclear fusion Other a. So I a r B...abandonment of all import controls , its findings on th: key problem of import dependence and security did not reflect a dear conviction that a

Regeneratively cooled coal combustor/gasifier with integral dry ash removal

DOEpatents

Beaufrere, A.H.

1982-04-30

A coal combustor/gasifier is disclosed which produces a low or medium combustion gas fired furnances or boilers. Two concentric shells define a combustion air flows to provide regenerative cooling of the inner shell for dry ash operation. A fuel flow and a combustion air flow having opposed swirls are mixed and burned in a mixing-combustion portion of the combustion volume and the ash laden combustion products flow with a residual swirl into an ash separation region. The ash is cooled below the fusion temperature and is moved to the wall by centrifugal force where it is entrained in the cool wall boundary layer. The boundary layer is stabilized against ash re-entrainment as it is moved to an ash removal annulus by a flow of air from the plenum through slots in the inner shell, and by suction on an ash removal skimmer slot.

Partial Cavity Flows at High Reynolds Numbers

NASA Astrophysics Data System (ADS)

Makiharju, Simo; Elbing, Brian; Wiggins, Andrew; Dowling, David; Perlin, Marc; Ceccio, Steven

2009-11-01

Partial cavity flows created for friction drag reduction were examined on a large-scale. Partial cavities were investigated at Reynolds numbers up to 120 million, and stable cavities with frictional drag reduction of more than 95% were attained at optimal conditions. The model used was a 3 m wide and 12 m long flat plate with a plenum on the bottom. To create the partial cavity, air was injected at the base of an 18 cm backwards-facing step 2.1 m from the leading edge. The geometry at the cavity closure was varied for different flow speeds to optimize the closure of the cavity. Cavity gas flux, thickness, frictional loads, and cavity pressures were measured over a range of flow speeds and air injection fluxes. High-speed video was used extensively to investigate the unsteady three dimensional cavity closure, the overall cavity shape and oscillations.

Induction melter apparatus

DOEpatents

Roach, Jay A [Idaho Falls, ID; Richardson, John G [Idaho Falls, ID; Raivo, Brian D [Idaho Falls, ID; Soelberg, Nicholas R [Idaho Falls, ID

2008-06-17

Apparatus and methods of operation are provided for a cold-crucible-induction melter for vitrifying waste wherein a single induction power supply may be used to effect a selected thermal distribution by independently energizing at least two inductors. Also, a bottom drain assembly may be heated by an inductor and may include an electrically resistive heater. The bottom drain assembly may be cooled to solidify molten material passing therethrough to prevent discharge of molten material therefrom. Configurations are provided wherein the induction flux skin depth substantially corresponds with the central longitudinal axis of the crucible. Further, the drain tube may be positioned within the induction flux skin depth in relation to material within the crucible or may be substantially aligned with a direction of flow of molten material within the crucible. An improved head design including four shells forming thermal radiation shields and at least two gas-cooled plenums is also disclosed.

Operating an induction melter apparatus

DOEpatents

Roach, Jay A.; Richardson, John G.; Raivo, Brian D.; Soelberg, Nicholas R.

2006-01-31

Apparatus and methods of operation are provided for a cold-crucible-induction melter for vitrifying waste wherein a single induction power supply may be used to effect a selected thermal distribution by independently energizing at least two inductors. Also, a bottom drain assembly may be heated by an inductor and may include an electrically resistive heater. The bottom drain assembly may be cooled to solidify molten material passing therethrough to prevent discharge of molten material therefrom. Configurations are provided wherein the induction flux skin depth substantially corresponds with the central longitudinal axis of the crucible. Further, the drain tube may be positioned within the induction flux skin depth in relation to material within the crucible or may be substantially aligned with a direction of flow of molten material within the crucible. An improved head design including four shells forming thermal radiation shields and at least two gas-cooled plenums is also disclosed.

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

1998-01-01

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

1998-06-30

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

PARFUME Theory and Model basis Report

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

Darrell L. Knudson; Gregory K Miller; G.K. Miller

2009-09-01

The success of gas reactors depends upon the safety and quality of the coated particle fuel. The fuel performance modeling code PARFUME simulates the mechanical, thermal and physico-chemical behavior of fuel particles during irradiation. This report documents the theory and material properties behind vari¬ous capabilities of the code, which include: 1) various options for calculating CO production and fission product gas release, 2) an analytical solution for stresses in the coating layers that accounts for irradiation-induced creep and swelling of the pyrocarbon layers, 3) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or amore » prismatic block core, as well as through the layers of each analyzed particle, 4) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, and kernel migration (or amoeba effect), 5) two independent methods for determining particle failure probabilities, 6) a model for calculating release-to-birth (R/B) ratios of gaseous fission products that accounts for particle failures and uranium contamination in the fuel matrix, and 7) the evaluation of an accident condition, where a particle experiences a sudden change in temperature following a period of normal irradiation. The accident condi¬tion entails diffusion of fission products through the particle coating layers and through the fuel matrix to the coolant boundary. This document represents the initial version of the PARFUME Theory and Model Basis Report. More detailed descriptions will be provided in future revisions.« less

Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

NASA Astrophysics Data System (ADS)

Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

2005-02-01

A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

Net energy payback and CO2 emissions from three midwestern wind farms: An update

USGS Publications Warehouse

White, S.W.

2006-01-01

This paper updates a life-cycle net energy analysis and carbon dioxide emissions analysis of three Midwestern utility-scale wind systems. Both the Energy Payback Ratio (EPR) and CO2 analysis results provide useful data for policy discussions regarding an efficient and low-carbon energy mix. The EPR is the amount of electrical energy produced for the lifetime of the power plant divided by the total amount of energy required to procure and transport the materials, build, operate, and decommission the power plants. The CO2 analysis for each power plant was calculated from the life-cycle energy input data. A previous study also analyzed coal and nuclear fission power plants. At the time of that study, two of the three wind systems had less than a full year of generation data to project the life-cycle energy production. This study updates the analysis of three wind systems with an additional four to eight years of operating data. The EPR for the utility-scale wind systems ranges from a low of 11 for a two-turbine system in Wisconsin to 28 for a 143-turbine system in southwestern Minnesota. The EPR is 11 for coal, 25 for fission with gas centrifuge enriched uranium and 7 for gaseous diffusion enriched uranium. The normalized CO2 emissions, in tonnes of CO2 per GW eh, ranges from 14 to 33 for the wind systems, 974 for coal, and 10 and 34 for nuclear fission using gas centrifuge and gaseous diffusion enriched uranium, respectively. ?? Springer Science+Business Media, LLC 2007.

A Preliminary Experimental Study of Vortex Tubes for Gas-Phase Fission Heating

DTIC Science & Technology

1959-02-20

CORPORATION for the U.S. ATOMIC ENERGY COMMISSION * RESTRICME DATA SECRET lo 710is a" me to so wavoied persea is pteW""Oo _____ -- -- --- LGAL OTICE... American , Canoga Park 40. AFPR, North American , Downey 41-42. Air Force Special Weapons Center 43. Air Research and Development Comand (RDTAPS) 44. Air

Energy: A Guide to Organizations and Information Resources in the United States.

ERIC Educational Resources Information Center

Center for California Public Affairs, Claremont.

A central source of information on the key organizations concerned with energy in the United States has been compiled. Chapter 2 covers organizations involved with broad questions of energy policy; Chapters 2-6 describe organizations having to do with sources of energy: oil, natural gas, coal, water power, nuclear fission, and alternate sources;…

SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

DOEpatents

Dickinson, R.W.

1963-03-01

This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

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

Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

Grain boundary (GB) separation as a mechanism for fission gas release (FGR), complementary to gas bubble interlinkage, has been experimentally observed in irradiated light water reactor fuel. However there has been limited effort to develop physics-based models incorporating this mechanism for the analysis of FGR. In this work, a computational study is carried out to investigate GB separation in UO2 fuel under the effect of gas bubble pressure and hydrostatic stress. A non-dimensional stress intensity factor formula is obtained through 2D axisymmetric analyses considering lenticular bubbles and Mode-I crack growth. The obtained functional form can be used in higher length-scalemore » models to estimate the contribution of GB separation to FGR.« less

Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

NASA Astrophysics Data System (ADS)

Van Dyke, Melissa; Martin, James

2005-02-01

The NASA Marshall Space Flight Center's Early Flight Fission Test Facility (EFF-TF), provides a facility to experimentally evaluate nuclear reactor related thermal hydraulic issues through the use of non-nuclear testing. This facility provides a cost effective method to evaluate concepts/designs and support mitigation of developmental risk. Electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004 which include the following. Initial evaluation of the Department of Energy Los Alamos National Laboratory 19 module stainless steel/sodium heat pipe reactor with integral gas heat exchanger was operated at up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37- pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to support future decisions regarding potential use of space nuclear systems for space exploration. All efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

Effect of rice husk ash mass on sustainability pyrolysis zone of fixed bed downdraft gasifier with capacity of 10 kg/hour

NASA Astrophysics Data System (ADS)

Surjosatyo, Adi; Haq, Imaduddin; Dafiqurrohman, Hafif; Gibran, Felly Rihlat

2017-03-01

The formation of pyrolysis sustainability (Sustainable Pyrolysis) is the objective of the gasification process. Pyrolysis zone in the gasification process is the result of the endothermic reaction that get heat from oxidation (combustion) of the fuel with oxygen, where cracking biomass rice husk result of such as charcoal, water vapor, steam tar, and gas - gas (CO, H 2, CH 4, CO 2 and N 2) and must be maintained at a pyrolysis temperature to obtain results plentiful gas (producer gas) or syngas (synthetic gas). Obtaining continuously syngas is indicated by flow rate (discharge) producer gas well and the consistency of the flame on the gas burner, it is highly influenced by the gasification process and the operation of the gasifier and the mass balance (mass balance) between the feeding rate of rice husk with the disposal of ash (ash removal). In experiments conducted is using fixed bed gasifier type downdraft capacity of 10 kg/h. Besides setting the mass of rice husks into the gasifier and disposal arrangements rice husk ash may affect the sustainability of the pyrolysis process, but tar produced during the gasification process causes sticky rice husk ash in the plenum gasifier. Modifications disposal system rice husk ash can facilitate the arrangement of ash disposal then could control the temperature pyrolysis with pyrolysis at temperatures between 500-750 ° C. The experimental study was conducted to determine the effect of mass quantities of rice husk ash issued against sustainability pyrolysis temperature which is obtained at each time disposal of rice husk ash to produce 60-90 grams of ash issued. From some experimental phenomena is expected to be seen pyrolysis and its effect on the flow rate of syngas and the stability of the flame on the gas burner so that this research can find a correlation to obtain performance (performance) gasifier optimal.

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

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

Hassan, Yassin; Anand, Nk

2016-03-30

A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensitymore » peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by comparing it with benchmark data. The experimental data gave the same pattern as the benchmark data. A turbulent single buoyant jet study was performed for the case of LOFC in the upper plenum of scaled VHTR. Time-averaged profiles show that 3,000 frames of images were sufficient for the study up to second-order statistics. Self-similarity is an important feature of jets since the behavior of jets is independent of Reynolds number and a sole function of geometry. Self-similarity profiles were well observed in the axial velocity and velocity magnitude profile regardless of z/D where the radial velocity did not show any similarity pattern. The normal components of Reynolds stresses have self-similarity within the expected range. The study shows that large vortices were observed close to the dome wall, indicating that the geometry of the VHTR has a significant impact on its safety and performance. Near the dome surface, large vortices were shown to inhibit the flows, resulting in reduced axial jet velocity. The vortices that develop subsequently reduce the Reynolds stresses that develop and the impact on the integrity of the VHTR upper plenum surface. Multiple jets study, including two, three and five jets, were investigated.« less

Comparative health and safety assessment of the SPS and alternative electrical generation systems

NASA Astrophysics Data System (ADS)

Habegger, L. J.; Gasper, J. R.; Brown, C. D.

1980-07-01

A comparative analysis of health and safety risks is presented for the Satellite Power System and five alternative baseload electrical generation systems: a low-Btu coal gasification system with an open-cycle gas turbine combined with a steam topping cycle; a light water fission reactor system without fuel reprocessing; a liquid metal fast breeder fission reactor system; a central station terrestrial photovoltaic system; and a first generation fusion system with magnetic confinement. For comparison, risk from a decentralized roof-top photovoltaic system with battery storage is also evaluated. Quantified estimates of public and occupational risks within ranges of uncertainty were developed for each phase of the energy system. The potential significance of related major health and safety issues that remain unquantitied are also discussed.

Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.; Coops, M.S.

1982-01-19

A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A nonoxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel.

Comparative health and safety assessment of the SPS and alternative electrical generation systems

NASA Technical Reports Server (NTRS)

Habegger, L. J.; Gasper, J. R.; Brown, C. D.

1980-01-01

A comparative analysis of health and safety risks is presented for the Satellite Power System and five alternative baseload electrical generation systems: a low-Btu coal gasification system with an open-cycle gas turbine combined with a steam topping cycle; a light water fission reactor system without fuel reprocessing; a liquid metal fast breeder fission reactor system; a central station terrestrial photovoltaic system; and a first generation fusion system with magnetic confinement. For comparison, risk from a decentralized roof-top photovoltaic system with battery storage is also evaluated. Quantified estimates of public and occupational risks within ranges of uncertainty were developed for each phase of the energy system. The potential significance of related major health and safety issues that remain unquantitied are also discussed.

Data summary report for fission product release test VI-6

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

Osborne, M.F.; Lorenz, R.A.; Travis, J.R.

Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of {approximately}42 MWd/kg, with inert gas release during irradiation of {approximately}2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directlymore » by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for {sup 85}Kr, 67% for {sup 129}I, 64% for {sup 125}Sb, 80% for both {sup 134}Cs and {sup 137}Cs, 14% for {sup 154}Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO{sub 2} and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model.« less

Compact multiwire proportional counters for the detection of fission fragments

NASA Astrophysics Data System (ADS)

Jhingan, Akhil; Sugathan, P.; Golda, K. S.; Singh, R. P.; Varughese, T.; Singh, Hardev; Behera, B. R.; Mandal, S. K.

2009-12-01

Two large area multistep position sensitive (two dimensional) multiwire proportional counters have been developed for experiments involving study of fission dynamics using general purpose scattering chamber facility at IUAC. Both detectors have an active area of 20×10 cm2 and provide position signals in horizontal (X) and vertical (Y) planes, timing signal for time of flight measurements and energy signal giving the differential energy loss in the active volume. The design features are optimized for the detection of low energy heavy ions at very low gas pressures. Special care was taken in setting up the readout electronics, constant fraction discriminators for position signals in particular, to get optimum position and timing resolutions along with high count rate handling capability of low energy heavy ions. A custom made charge sensitive preamplifier, having lower gain and shorter decay time, has been developed for extracting the differential energy loss signal. The position and time resolutions of the detectors were determined to be 1.1 mm full width at half maximum (FWHM) and 1.7 ns FWHM, respectively. The detector could handle heavy ion count rates exceeding 20 kHz without any breakdown. Time of flight signal in combination with differential energy loss signal gives a clean separation of fission fragments from projectile and target like particles. The timing and position signals of the detectors are used for fission coincidence measurements and subsequent extraction of their mass, angular, and total kinetic energy distributions. This article describes systematic study of these fission counters in terms of efficiency, time resolution, count rate handling capability, position resolution, and the readout electronics. The detector has been operated with both five electrode geometry and four electrode geometry, and a comparison has been made in their performances.

A 3-D turbulent flow analysis using finite elements with k-ɛ model

NASA Astrophysics Data System (ADS)

Okuda, H.; Yagawa, G.; Eguchi, Y.

1989-03-01

This paper describes the finite element turbulent flow analysis, which is suitable for three-dimensional large scale problems. The k-ɛ turbulence model as well as the conservation equations of mass and momentum are discretized in space using rather low order elements. Resulting coefficient matrices are evaluated by one-point quadrature in order to reduce the computational storage and the CPU cost. The time integration scheme based on the velocity correction method is employed to obtain steady state solutions. For the verification of this FEM program, two-dimensional plenum flow is simulated and compared with experiment. As the application to three-dimensional practical problems, the turbulent flows in the upper plenum of the fast breeder reactor are calculated for various boundary conditions.

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1961-01-24

A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

Station blackout at Browns Ferry Unit One: iodine and noble-gas distribution and release

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

Wichner, R.P.; Weber, C.F.; Lorenz, R.A.

1982-08-01

This is the second volume of a report describing the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to a postulated Station Blackout, defined as a loss of offsite power combined with failure of all onsite emergency diesel-generators to start and load. The Station Blackout is assumed to persist beyond the point of battery exhaustion and the completely powerless state leads to core uncovery, meltdown, reactor vessel failure, and failure of the primary containment by overtemperature-induced degradation of the electrical penetration assembly seals. The sequence of events is described in Volume 1; the material in this volumemore » deals with the analysis of fission product noble gas and iodine transport during the accident. Factors which affect the fission product movements through the series of containment design barriers are reviewed. For a reactive material such as iodine, proper assessment of the rate of movement requires determination of the chemical changes along the pathway which alter the physical properties such as vapor pressure and solubility and thereby affect the transport rate. A methodology for accomplishing this is demonstrated in this report.« less

A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

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

Petti, David Andrew

2017-04-01

Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however,more » is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.« less

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

DOE PAGES

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; ...

2015-09-03

Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

Measurement of activation of helium gas by 238U beam irradiation at about 11 A MeV

NASA Astrophysics Data System (ADS)

Akashio, A.; Tanaka, K.; Imao, H.; Uwamino, Y.

2017-09-01

A new helium-gas stripper system has been applied at the 11 A MeV uranium beam of the Radioactive Isotope Beam Factory of the RIKEN accelerator facility. Although the gas stripper is important for the heavy-ion accelerator facility, the residual radiation that is generated is a serious problem for maintenance work. The residual dose was evaluated by using three-layered activation samples of aluminium and bismuth. The γ-rays from produced radionuclides with in-flight fission of the 238U beam and from the material of the chamber activated by neutrons were observed by using a Ge detector and compared with the values calculated by using the Monte-Carlo simulation code PHITS.

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Gaseous-fuel nuclear reactor research for multimegawatt power in space

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Towards saturation of the electron-capture delayed fission probability: The new isotopes 240Es and 236Bk

NASA Astrophysics Data System (ADS)

Konki, J.; Khuyagbaatar, J.; Uusitalo, J.; Greenlees, P. T.; Auranen, K.; Badran, H.; Block, M.; Briselet, R.; Cox, D. M.; Dasgupta, M.; Di Nitto, A.; Düllmann, Ch. E.; Grahn, T.; Hauschild, K.; Herzán, A.; Herzberg, R.-D.; Heßberger, F. P.; Hinde, D. J.; Julin, R.; Juutinen, S.; Jäger, E.; Kindler, B.; Krier, J.; Leino, M.; Lommel, B.; Lopez-Martens, A.; Luong, D. H.; Mallaburn, M.; Nishio, K.; Pakarinen, J.; Papadakis, P.; Partanen, J.; Peura, P.; Rahkila, P.; Rezynkina, K.; Ruotsalainen, P.; Sandzelius, M.; Sarén, J.; Scholey, C.; Sorri, J.; Stolze, S.; Sulignano, B.; Theisen, Ch.; Ward, A.; Yakushev, A.; Yakusheva, V.

2017-01-01

The new neutron-deficient nuclei 240Es and 236Bk were synthesised at the gas-filled recoil separator RITU. They were identified by their radioactive decay chains starting from 240Es produced in the fusion-evaporation reaction 209Bi(34S,3n)240Es. Half-lives of 6 (2)s and 22-6+13s were obtained for 240Es and 236Bk, respectively. Two groups of α particles with energies Eα = 8.19 (3)MeV and 8.09 (3)MeV were unambiguously assigned to 240Es. Electron-capture delayed fission branches with probabilities of 0.16 (6) and 0.04 (2) were measured for 240Es and 236Bk, respectively. These new data show a continuation of the exponential increase of ECDF probabilities in more neutron-deficient isotopes.

Fission product palladium-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

1990-07-01

Interaction of fission product palladium (Pd) with the silicon carbide (SiC) layer was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors (HTGR) with an optical microscope and electron probe microanalyzers. The SiC layers were attacked locally or the reaction product formed nodules at the attack site. Although the main element concerned with the reaction was palladium, rhodium and ruthenium were also detected at the corroded areas in some particles. Palladium was detected on both the hot and cold sides of the particles, but the corroded areas and the palladium accumulations were distributed particularly on the cold side of the particles. The observed Pd-SiC reaction depths were analyzed on the assumption that the release of palladium from the fuel kernel controls the whole Pd-SiC reaction.

Nuclear reactor flow control method and apparatus

DOEpatents

Church, J.P.

1993-03-30

Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Nuclear reactor flow control method and apparatus

DOEpatents

Church, John P.

1993-01-01

Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Flexible ceramic gasket for SOFC generator

DOEpatents

Zafred, Paolo [Murrysville, PA; Prevish, Thomas [Trafford, PA

2009-02-03

A solid oxide fuel cell generator (10) contains stacks of hollow axially elongated fuel cells (36) having an open top end (37), an oxidant inlet plenum (52), a feed fuel plenum (11), a combustion chamber (94) for combusting reacted oxidant/spent fuel; and, optionally, a fuel recirculation chamber (106) below the combustion chamber (94), where the fuel recirculation chamber (94) is in part defined by semi-porous fuel cell positioning gasket (108), all within an outer generator enclosure (8), wherein the fuel cell gasket (108) has a laminate structure comprising at least a compliant fibrous mat support layer and a strong, yet flexible woven layer, which may contain catalytic particles facing the combustion chamber, where the catalyst, if used, is effective to further oxidize exhaust fuel and protect the open top end (37) of the fuel cells.

Transient Fission Gas Behavior in Uranium Nitride Fuel Under Proposed Space Applications

DTIC Science & Technology

1991-12-01

Press, Oxford UK, 1959). [7] M.H. Wood, J. Nucl. Mater. 82 (1979) 264 . [8] H. Blank and Hj. Matzke, Rad. Effects. 17 (1973) 57. [9] C. Ronchi and P.T...Benson, P.I. Freeman and E. Dempsey, J. Am. Ceramic Soc. 46 (1963) 43. [77] H.R. Warner and F.A. Nichols, Report WAPD -T-2161 (Bettis Atomic Power

A METHOD OF PREPARING URANIUM DIOXIDE

DOEpatents

Scott, F.A.; Mudge, L.K.

1963-12-17

A process of purifying raw, in particular plutonium- and fission- products-containing, uranium dioxide is described. The uranium dioxide is dissolved in a molten chloride mixture containing potassium chloride plus sodium, lithium, magnesium, or lead chloride under anhydrous conditions; an electric current and a chlorinating gas are passed through the mixture whereby pure uranium dioxide is deposited on and at the same time partially redissolved from the cathode. (AEC)

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

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

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

An Assessment of NASA Glenn's Aeroacoustic Experimental and Predictive Capabilities for Installed Cooling Fans. Part 1; Aerodynamic Performance

NASA Technical Reports Server (NTRS)

VanZante, Dale E.; Koch, L. Danielle; Wernet, Mark P.; Podboy, Gary G.

2006-01-01

Driven by the need for low production costs, electronics cooling fans have evolved differently than the bladed components of gas turbine engines which incorporate multiple technologies to enhance performance and durability while reducing noise emissions. Drawing upon NASA Glenn's experience in the measurement and prediction of gas turbine engine aeroacoustic performance, tests have been conducted to determine if these tools and techniques can be extended for application to the aerodynamics and acoustics of electronics cooling fans. An automated fan plenum installed in NASA Glenn's Acoustical Testing Laboratory was used to map the overall aerodynamic and acoustic performance of a spaceflight qualified 80 mm diameter axial cooling fan. In order to more accurately identify noise sources, diagnose performance limiting aerodynamic deficiencies, and validate noise prediction codes, additional aerodynamic measurements were recorded for two operating points: free delivery and a mild stall condition. Non-uniformities in the fan s inlet and exhaust regions captured by Particle Image Velocimetry measurements, and rotor blade wakes characterized by hot wire anemometry measurements provide some assessment of the fan aerodynamic performance. The data can be used to identify fan installation/design changes which could enlarge the stable operating region for the fan and improve its aerodynamic performance and reduce noise emissions.

Advanced Duct Sealing Testing

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

Sherman, Max H.; Walker, Iain S.

Duct leakage has been identified as a major source of energy loss in residential buildings. Most duct leakage occurs at the connections to registers, plenums or branches in the duct system. At each of these connections a method of sealing the duct system is required. Typical sealing methods include tapes or mastics applied around the joints in the system. Field examinations of duct systems have typically shown that these seals tend to fail over extended periods of time. The Lawrence Berkeley National Laboratory has been testing sealant durability for several years. Typical duct tape (i.e. fabric backed tapes with naturalmore » rubber adhesives) was found to fail more rapidly than all other duct sealants. This report summarizes the results of duct sealant durability testing of five UL 181B-FX listed duct tapes (three cloth tapes, a foil tape and an Oriented Polypropylene (OPP) tape). One of the cloth tapes was specifically developed in collaboration with a tape manufacturer to perform better in our durability testing. The first test involved the aging of common ''core-to-collar joints'' of flexible duct to sheet metal collars, and sheet metal ''collar-to-plenum joints'' pressurized with 200 F (93 C) air. The second test consisted of baking duct tape specimens in a constant 212 F (100 C) oven following the UL 181B-FX ''Temperature Test'' requirements. Additional tests were also performed on only two tapes using sheet metal collar-to-plenum joints. Since an unsealed flexible duct joint can have a variable leakage depending on the positioning of the flexible duct core, the durability of the flexible duct joints could not be based on the 10% of unsealed leakage criteria. Nevertheless, the leakage of the sealed specimens prior to testing could be considered as a basis for a failure criteria. Visual inspection was also documented throughout the tests. The flexible duct core-to-collar joints were inspected monthly, while the sheet metal collar-to-plenum joints were inspected weekly. The baking test specimens were visually inspected weekly, and the durability was judged by the observed deterioration in terms of brittleness, cracking, flaking and blistering (the terminology used in the UL 181B-FX test procedure).« less

Use of double and triple-ion irradiation to study the influence of high levels of helium and hydrogen on void swelling of 8-12% Cr ferritic-martensitic steels

NASA Astrophysics Data System (ADS)

Kupriiyanova, Y. E.; Bryk, V. V.; Borodin, O. V.; Kalchenko, A. S.; Voyevodin, V. N.; Tolstolutskaya, G. D.; Garner, F. A.

2016-01-01

In accelerator-driven spallation (ADS) devices, some of the structural materials will be exposed to intense fluxes of very high energy protons and neutrons, producing not only displacement damage, but very high levels of helium and hydrogen. Unlike fission flux-spectra where most helium and hydrogen are generated by transmutation in nickel and only secondarily in iron or chromium, gas production in ADS flux-spectra are rather insensitive to alloy composition, such that Fe-Cr base ferritic alloys also generate very large gas levels. While ferritic alloys are known to swell less than austenitic alloys in fission spectra, there is a concern that high gas levels in fusion and especially ADS facilities may strongly accelerate void swelling in ferritic alloys. In this study of void swelling in response to helium and hydrogen generation, irradiation was conducted on three ferritic-martensitic steels using the Electrostatic Accelerator with External Injector (ESUVI) facility that can easily produce any combination of helium to dpa and/or hydrogen to dpa ratios. Irradiation was conducted under single, dual and triple beam modes using 1.8 MeV Cr+3, 40 keV He+, and 20 keV H+. In the first part of this study we investigated the response of dual-phase EP-450 to variations in He/dpa and H/dpa ratio, focusing first on dual ion studies and then triple ion studies, showing that there is a diminishing influence on swelling with increasing total gas content. In the second part we investigated the relative response of three alloys spanning a range of starting microstructure and composition. In addition to observing various synergisms between He and H, the most important conclusion was that the tempered martensite phase, known to lag behind the ferrite phase in swelling in the absence of gases, loses much of its resistance to void nucleation when irradiated at large gas/dpa levels.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Daniels, F.

1957-10-15

Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

OPERATIONAL CHARACTERISTICS OF THE ARMOUR FISSION GAS GAMMA FACILITY

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

Terrell, C.W.; McElroy, W.N.

1958-10-31

As the reactor power level is changed frequently, the radiation levels in the gamma facility fluctuate. Data are presented to show the power dependency of the gamma dose rate and the manner of growth and decay. Additional data show the dependercy of the equilibrium gamma activity on the foel temperature and total system pressure. The final phase of the work is directed toward determining an average gamma energy by attenuation measurements with various thicknesses of several materials. The neutrou flux associated with the gas phase activity is determined by foil measurement. From the measurements of dose rate and average gammamore » energy, calculations to determine the number of curies of gas phase decay gamma activity per watt of reactor power are presented. (auth)« less

Lognormal field size distributions as a consequence of economic truncation

USGS Publications Warehouse

Attanasi, E.D.; Drew, L.J.

1985-01-01

The assumption of lognormal (parent) field size distributions has for a long time been applied to resource appraisal and evaluation of exploration strategy by the petroleum industry. However, frequency distributions estimated with observed data and used to justify this hypotheses are conditional. Examination of various observed field size distributions across basins and over time shows that such distributions should be regarded as the end result of an economic filtering process. Commercial discoveries depend on oil and gas prices and field development costs. Some new fields are eliminated due to location, depths, or water depths. This filtering process is called economic truncation. Economic truncation may occur when predictions of a discovery process are passed through an economic appraisal model. We demonstrate that (1) economic resource appraisals, (2) forecasts of levels of petroleum industry activity, and (3) expected benefits of developing and implementing cost reducing technology are sensitive to assumptions made about the nature of that portion of (parent) field size distribution subject to economic truncation. ?? 1985 Plenum Publishing Corporation.

Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

NASA Astrophysics Data System (ADS)

Barber, Duncan Henry

During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A benchmark calculation demonstrates the improvement in agreement of the total inventory of those chemical elements included in the RMC fuel model to an ORIGEN-S calculation. ORIGEN-S is the Oak Ridge isotope generation and depletion computer program. The Gibbs energy minimizer requires a chemical database containing coefficients from which the Gibbs energy of pure compounds, gas and liquid mixtures, and solid solutions can be calculated. The RMC model of irradiated uranium dioxide fuel has been converted into the required format. The Gibbs energy minimizer has been incorporated into a new model of fission-product vaporization from the fuel surface. Calculated release fractions using the new code have been compared to results calculated with SOURCE IST 2.0P11 and to results of tests used in the validation of SOURCE 2.0. The new code shows improvements in agreement with experimental releases for a number of nuclides. Of particular significance is the better agreement between experimental and calculated release fractions for 140La. The improved agreement reflects the inclusion in the RMC model of the solubility of lanthanum (III) oxide (La2O3) in the fuel matrix. Calculated lanthanide release fractions from earlier computer programs were a challenge to environmental qualification analysis of equipment for some accident scenarios. The new prototype computer program would alleviate this concern. Keywords: Nuclear Engineering; Material Science; Thermodynamics; Radioactive Material, Gibbs Energy Minimization, Actinide Generation and Depletion, FissionProduct Generation and Depletion.

40 CFR 63.10906 - What definitions apply to this subpart?

Code of Federal Regulations, 2010 CFR

2010-07-01

..., plenums, and fans. Chlorinated plastics means solid polymeric materials that contain chlorine in the polymer chain, such as polyvinyl chloride (PVC) and PVC copolymers. Control device means the air pollution...

40 CFR 63.10906 - What definitions apply to this subpart?

Code of Federal Regulations, 2011 CFR

2011-07-01

..., plenums, and fans. Chlorinated plastics means solid polymeric materials that contain chlorine in the polymer chain, such as polyvinyl chloride (PVC) and PVC copolymers. Control device means the air pollution...

System for supporting a bundled tube fuel injector within a combustor

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

LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold

A combustor includes an end cover having an outer side and an inner side, an outer barrel having a forward end that is adjacent to the inner side of the end cover and an aft end that is axially spaced from the forward end. An inner barrel is at least partially disposed concentrically within the outer barrel and is fixedly connected to the outer barrel. A fluid conduit extends downstream from the end cover. A first bundled tube fuel injector segment is disposed concentrically within the inner barrel. The bundled tube fuel injector segment includes a fuel plenum that ismore » in fluid communication with the fluid conduit and a plurality of parallel tubes that extend axially through the fuel plenum. The bundled tube fuel injector segment is fixedly connected to the inner barrel.« less

Water jacket for solid particle solar receiver

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

Wasyluk, David T.

A solar receiver includes: water jacket panels each having a light-receiving side and a back side with a watertight sealed plenum defined in-between; light apertures passing through the watertight sealed plenums to receive light from the light-receiving sides of the water jacket panels; a heat transfer medium gap defined between the back sides of the water jacket panels and a cylindrical back plate; and light channeling tubes optically coupled with the light apertures and extending into the heat transfer medium gap. In some embodiments ends of the light apertures at the light receiving side of the water jacket panel aremore » welded together to define at least a portion of the light-receiving side. A cylindrical solar receiver may be constructed using a plurality of such water jacket panels arranged with their light-receiving sides facing outward.« less

Flow and Temperature Distribution Evaluation on Sodium Heated Large-sized Straight Double-wall-tube Steam Generator

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

Kisohara, Naoyuki; Moribe, Takeshi; Sakai, Takaaki

2006-07-01

The sodium heated steam generator (SG) being designed in the feasibility study on commercialized fast reactor cycle systems is a straight double-wall-tube type. The SG is large sized to reduce its manufacturing cost by economics of scale. This paper addresses the temperature and flow multi-dimensional distributions at steady state to obtain the prospect of the SG. Large-sized heat exchanger components are prone to have non-uniform flow and temperature distributions. These phenomena might lead to tube buckling or tube to tube-sheet junction failure in straight tube type SGs, owing to tubes thermal expansion difference. The flow adjustment devices installed in themore » SG are optimized to prevent these issues, and the temperature distribution properties are uncovered by analysis methods. The analysis model of the SG consists of two parts, a sodium inlet distribution plenum (the plenum) and a heat transfer tubes bundle region (the bundle). The flow and temperature distributions in the plenum and the bundle are evaluated by the three-dimensional code 'FLUENT' and the two dimensional thermal-hydraulic code 'MSG', respectively. The MSG code is particularly developed for sodium heated SGs in JAEA. These codes have revealed that the sodium flow is distributed uniformly by the flow adjustment devices, and that the lateral tube temperature distributions remain within the allowable temperature range for the structural integrity of the tubes and the tube to tube-sheet junctions. (authors)« less

Mockup Small-Diameter Air Distribution System

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

A. Poerschke and A. Rudd

2016-05-01

This report investigates the feasibility of using a home-run manifold small-diameter duct system to provide space conditioning air to individual thermal zones in a low-load home. This compact layout allows duct systems to be brought easily within conditioned space via interior partition walls. Centrally locating the air hander unit in the house significantly reduces duct lengths. The plenum box is designed so that each connected duct receives an equal amount of airflow, regardless of the duct position on the box. Furthermore, within a reasonable set of length restrictions, each duct continues to receive similar airflow. The design method uses anmore » additive approach to reach the total needed zonal airflow. Once the cubic feet per minute needed to satisfy the thermal load of a zone has been determined, the total number of duct runs to a zone can be calculated by dividing the required airflow by the standard airflow from each duct. The additive approach greatly simplifies the design effort and reduces the potential for duct design mistakes to be made. Measured results indicate that this plenum design can satisfy the heating load. However, the total airflow falls short of satisfying the cooling load in a hypothetical building. Static pressure inside the plenum box of 51.5 Pa limited the total airflow of the attached mini-split heat pump blower, thus limiting the total thermal capacity. Fan energy consumption is kept to 0.16 to 0.22 watt/CFM by using short duct runs and smooth duct material.« less

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

This report investigates the feasibility of using a home-run manifold small-diameter duct system to provide space conditioning air to individual thermal zones in a low-load home. This compact layout allows duct systems to be brought easily within conditioned space via interior partition walls. Centrally locating the air hander unit in the house significantly reduces duct lengths. The plenum box is designed so that each connected duct receives an equal amount of airflow, regardless of the duct position on the box. Furthermore, within a reasonable set of length restrictions, each duct continues to receive similar airflow. The design method uses anmore » additive approach to reach the total needed zonal airflow. Once the cubic feet per minute needed to satisfy the thermal load of a zone has been determined, the total number of duct runs to a zone can be calculated by dividing the required airflow by the standard airflow from each duct. The additive approach greatly simplifies the design effort and reduces the potential for duct design mistakes to be made. Measured results indicate that this plenum design can satisfy the heating load. However, the total airflow falls short of satisfying the cooling load in a hypothetical building. Static pressure inside the plenum box of 51.5 Pa limited the total airflow of the attached mini-split heat pump blower, thus limiting the total thermal capacity. Fan energy consumption is kept to 0.16 to 0.22 watt/CFM by using short duct runs and smooth duct material.« less

Advanced nuclear thermal propulsion concepts

NASA Technical Reports Server (NTRS)

Howe, Steven D.

1993-01-01

In 1989, a Presidential directive created the Space Exploration Initiative (SEI) which had a goal of placing mankind on Mars in the early 21st century. The SEI was effectively terminated in 1992 with the election of a new administration. Although the initiative did not exist long enough to allow substantial technology development, it did provide a venue, for the first time in 20 years, to comprehensively evaluate advanced propulsion concepts which could enable fast, manned transits to Mars. As part of the SEI based investigations, scientists from NASA, DoE National Laboratories, universities, and industry met regularly and proceeded to examine a variety of innovative ideas. Most of the effort was directed toward developing a solid-core, nuclear thermal rocket and examining a high-power nuclear electric propulsion system. In addition, however, an Innovative Concepts committee was formed and charged with evaluating concepts that offered a much higher performance but were less technologically mature. The committee considered several concepts and eventually recommended that further work be performed in the areas of gas core fission rockets, inertial confinement fusion systems, antimatter based rockets, and gas core fission electric systems. Following the committee's recommendations, some computational modeling work has been performed at Los Alamos in certain of these areas and critical issues have been identified.

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

NASA Astrophysics Data System (ADS)

Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Examination of UC-ZrC after long term irradiation at thermionic temperature

NASA Technical Reports Server (NTRS)

Yang, L.; Johnson, H. O.

1972-01-01

Two fluoride tungsten clad UC-ZrC fueled capsules, designated as V-2C and V-2D, were examined a hot cell after irradiation in NASA Plum Brook Reactor at a maximum cladding temperature of 1930 K for 11,089 and 12,031 hours to burnups of 3.0 x 10 to the 20th power and 2.1 x 10 to the 20th power fission/c.c. respectively. Percentage of fission gas release from the fuel material was measured by radiochemical means. Cladding deformation, fuel-cladding interaction and microstructures of fuel, cladding, and fuel-cladding interface were studied metallographically. Compositions of dispersions in fuel, fuel matrix and fuel-cladding interaction layer were analyzed by electron microprobe techniques. Axial and radial distributions of burnup were determined by gamma-scan, autoradiography and isotopic burnup analysis. The results are presented and discussed in conjunction with the requirements of thermionic fuel elements for space power application.

Simulation of xenon, uranium vacancy and interstitial diffusion and grain boundary segregation in UO 2

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

Andersson, Anders D.; Tonks, Michael R.; Casillas, Luis

2014-10-31

In light water reactor fuel, gaseous fission products segregate to grain boundaries, resulting in the nucleation and growth of large intergranular fission gas bubbles. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations 1, continuum models for diffusion of xenon (Xe), uranium (U) vacancies and U interstitials in UO 2 have been derived for both intrinsic conditions and under irradiation. Segregation of Xe to grain boundaries is described by combining the bulk diffusion model with a model for the interaction between Xe atoms and three different grain boundaries in UO 2 ( Σ5 tilt, Σ5more » twist and a high angle random boundary),as derived from atomistic calculations. All models are implemented in the MARMOT phase field code, which is used to calculate effective Xe and U diffusivities as well as redistribution for a few simple microstructures.« less

SBLOCA outside containment at Browns Ferry Unit One. Volume 2. Iodine, cesium, and noble gas distribution and release

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

Wichner, R.P.; Weber, C.F.; Wright, A.L.

1983-09-01

This is the second volume of a two-part study regarding the response of Browns Ferry Unit 1 to a postulated break in the scram discharge volume of the control rod drive hydraulic system immediately following a scram. The material in this second volume pertains to the second aspect of the study, the resultant transport of fission products from their original locations in the fuel to a series of repositories within the primary system, the primary and secondary containment structures, and ultimately the release of a small portion to the environment. Transport models are developed for the noble gases krypton andmore » xenon and for iodine and cesium to describe the release of these fission products from the overheated fuel and their subsequent movement under the conditions predicted to exist in the various repositories during the course of the accident.« less

Precise ruthenium fission product isotopic analysis using dynamic reaction cell inductively coupled plasma mass spectrometry (DRC-ICP-MS)

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

Brown, Christopher F.; Dresel, P. Evan; Geiszler, Keith N.

2006-05-09

99Tc is a subsurface contaminant of interest at numerous federal, industrial, and international facilities. However, as a mono-isotopic fission product, 99Tc lacks the ability to be used as a signature to differentiate between the different waste disposal pathways that could have contributed to subsurface contamination at these facilities. Ruthenium fission-product isotopes are attractive analogues for the characterization of 99Tc sources because of their direct similarity to technetium with regard to subsurface mobility, and their large fission yields and low natural background concentrations. We developed an inductively coupled plasma mass spectrometry (ICP-MS) method capable of measuring ruthenium isotopes in groundwater samplesmore » and extracts of vadose zone sediments. Samples were analyzed directly on a Perkin Elmer ELAN DRC II ICP-MS after a single pass through a 1-ml bed volume of Dowex AG 50W-X8 100-200 mesh cation exchange resin. Precise ruthenium isotopic ratio measurements were achieved using a low-flow Meinhard-type nebulizer and long sample acquisition times (150,000 ms). Relative standard deviations of triplicate replicates were maintained at less than 0.5% when the total ruthenium solution concentration was 0.1 ng/ml or higher. Further work was performed to minimize the impact caused by mass interferences using the dynamic reaction cell (DRC) with O2 as the reaction gas. The aqueous concentrations of 96Mo and 96Zr were reduced by more than 99.7% in the reaction cell prior to injection of the sample into the mass analyzer quadrupole. The DRC was used in combination with stable-mass correction to quantitatively analyze samples containing up to 2-orders of magnitude more zirconium and molybdenum than ruthenium. The analytical approach documented herein provides an efficient and cost-effective way to precisely measure ruthenium isotopes and quantitate total ruthenium (natural vs. fission-product) in aqueous matrixes.« less

Early evolution of Martian volatiles: Nitrogen and noble gas components in ALH84001 and Chassigny

NASA Astrophysics Data System (ADS)

Mathew, K. J.; Marti, K.

2001-01-01

Studies on SNC meteorites have permitted the characterization of modern Martian atmospheric components as well as indigenous Martian nitrogen and solar-type xenon. New isotopic and elemental abundances of noble gases and nitrogen in ALH84001 and Chassigny provide important constraints on the early evolution of the planet. A primitive solar Xe component (Chass-S) and an evolved Xe component (Chass-E), augmented with fission Xe are identified in Chassigny. Both components represent interior reservoirs of Mars and are characterized by low 129Xe/132Xe (<1.07) and by distinct elemental ratios 36Ar/132Xe<5 and >130, respectively. Light nitrogen (δ15N=-30‰) is associated with the Chass-S component and is enriched in melt inclusions in olivine. An ancient (presumably incorporated ~4 Gyr ago) evolved Martian atmospheric component is identified in ALH84001 and has the following signatures: 129Xe/132Xe=2.16, 36Ar/38Ar>=5.0, 36Ar/132Xe=~50, 84Kr/132Xe=~6, and δ15N=7‰. The trapped Xe component in ALH84001 is not isotopically fractionated. We observe major shifts in nitrogen signatures due to cosmogenic N component in both Chassigny and ALH84001. A heavy nitrogen component of comparable magnitude (δ15N>150‰) has previously been interpreted as (heavy) Martian atmospheric N. In situ produced fission Xe components, due to 244Pu in ALH84001 and due to 238U in Chassigny, are identified. The ALH84001 data strongly constrain exchanges of Martian atmospheric and interior reservoirs. Mars retained abundant fission Xe components, and this may account for the low observed fission Xe component in the modern Martian atmosphere. Chronometric information regarding the evolution of the early Martian atmosphere can be secured from the relative abundances of radiogenic and fission Xe, as ~80% of the Martian 129Xer is observed in the atmospheric 129Xe/132Xe ratio ~ 4 Gyr ago.

24 CFR 3280.715 - Circulating air systems.

Code of Federal Regulations, 2011 CFR

2011-04-01

... other listed air-cooler coil is installed between the furnace and the duct plenum, the total static... applied only to surfaces that are dry and dust-, dirt-, oil-, and grease-free. Tapes and mastic closure...

24 CFR 3280.715 - Circulating air systems.

Code of Federal Regulations, 2014 CFR

2014-04-01

... listed air-cooler coil is installed between the furnace and the duct plenum, the total static pressure... applied only to surfaces that are dry and dust-, dirt-, oil-, and grease-free. Tapes and mastic closure...

24 CFR 3280.715 - Circulating air systems.

Code of Federal Regulations, 2012 CFR

2012-04-01

... other listed air-cooler coil is installed between the furnace and the duct plenum, the total static... applied only to surfaces that are dry and dust-, dirt-, oil-, and grease-free. Tapes and mastic closure...

24 CFR 3280.715 - Circulating air systems.

Code of Federal Regulations, 2010 CFR

2010-04-01

... other listed air-cooler coil is installed between the furnace and the duct plenum, the total static... applied only to surfaces that are dry and dust-, dirt-, oil-, and grease-free. Tapes and mastic closure...

24 CFR 3280.715 - Circulating air systems.

Code of Federal Regulations, 2013 CFR

2013-04-01

... other listed air-cooler coil is installed between the furnace and the duct plenum, the total static... applied only to surfaces that are dry and dust-, dirt-, oil-, and grease-free. Tapes and mastic closure...

FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF BUBBLER CONFIGURATIONS USING HLW AZ-101 SIMULANTS VSL-04R4800-4 REV 0 10/5/04

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

KRUGER AA; MATLACK KS; GONG W

2011-12-29

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on themore » results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was achieved, which was used as an indicator of a maximized feed rate for each test. The first day of each test was used to build the cold cap and decrease the plenum temperature. The remainder of each test was split into two- to six-day segments, each with a different bubbling rate, bubbler orientation, or feed concentration of chloride and sulfur.« less

Nuclear propulsion - A vital technology for the exploration of Mars and the planets beyond

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1989-01-01

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the only other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulse of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class spaceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.

Gamma Ray Spectroscopy and SASSYER

NASA Astrophysics Data System (ADS)

Pauerstein, Benjamin; Bonniwell, Cain; Allmond, J. M.; Beausang, C. W.

2009-10-01

An experiment was performed to study the Gd and Tb nuclei resulting from a 27 MeV proton beam on a 156Gd target. This was conducted at Lawrence Berkeley National Laboratory using the STARS/LIBERACE array. The main focus of the experiment was on charged particle channels (p,d) into 155Gd and (p,t) into 154Gd. However, the trigger was either gamma-gamma or particle-gamma so new data was also obtained on 155Tb nuclei following fusion evaporation reactions. Preliminary analysis was conducted at Wright Nuclear Structure Lab where RADWARE programs were used to analyze the data and search for unknown gamma rays. A second, separate, experiment was conducted using the SASSYER (a gas-filled separator at Yale). In this experiment, fission fragments from a 252Cf source were focused to a DSSD and a Ge detector was used to search for either gamma-decay from long lived isomers in the fission fragments or to find gammas from recoil-beta-decay tagging on the fission fragments. The data collection seems to have gone smoothly, and the data is currently being sorted for analysis. This work was supported by the US Department of Energy under grant numbers DE-FG02-52NA26206 and DE-FG02-05ER41379.

Nuclear propulsion: a vital technology for the exploration of Mars and the planets beyond

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

Borowski, S.K.

1988-01-01

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the olny other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-Earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulsemore » of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class speceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.« less

Nuclear propulsion: A vital technology for the exploration of Mars and the planets beyond

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1988-01-01

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the olny other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-Earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulse of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class speceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.

Evidence of extinct 244Pu in ancient terrestrial zircons

NASA Astrophysics Data System (ADS)

Harrison, T. M.; Turner, G.; Holland, G.; Gilmour, J. D.; Mojzsis, S. J.

2003-04-01

The Pu/U ratio of the early Earth is an important parameter in models of mantle evolution based on noble gas isotopes. Current estimates assume the Earth accreted with a chondritic Pu/U and are based on analyses of the chondrite St Severin and the achondrite Angra dos Reis. These estimates are poorly constrained, ranging from 0.004 to 0.008. On account of its short, 82 Ma, half-life, 244Pu was essentially extinct 3,900 Ma ago, and consequently there exists no reliable measurement of Pu/U for the Earth. The discovery of zircons dating from the period when 244Pu was "live" offers the first opportunity to measure the former terrestrial abundance of 244Pu directly. Xenon isotopes are produced by spontaneous fission and, in principle, are readily distinguishable from those produced by 238U-fission (e.g. 131Xe/136Xe = 0.24 and 0.08 respectively). However the expected levels of fission xenon in individual zircons, weighing 1 - 2 μg and containing 100 - 200 ppm U, are below, or at best comparable to, the Xe blank levels (˜10-15 ccSTP) typical of conventional noble gas mass spectrometers. In order to analyse these minute amounts of xenon we have made use of a uniquely sensitive instrument, developed in Manchester, based on the principle of laser resonance ionisation. RELAX (Refrigerator Enhanced Laser Analyser for Xenon) is capable of analysing samples of only a few thousand atoms, some two orders of magnitude smaller than conventional mass spectrometers. We have carried out preliminary analyses of 4 individual 4,150 Ma zircons and one 3,600 Ma zircon from Jack Hills, Western Australia, and obtained five clear fission spectra. All but one were essentially free from significant atmospheric blank (the average 130Xe blank was 3× 10-18 ccSTP, i.e. 80 atoms). The spectra of the older zircons clearly demonstrated the presence of varying amounts of 244Pu fission xenon. The highest 131Xe/136Xe, 0.136 ± 0.003, corresponds to an initial Pu/U ratio of 0.0057. The lower ratios could result from loss of Xe after 4.0 Ga or represent U-Pu fractionation. We are currently repeating the exercise with older zircons and intend to search for correlations with REE patterns, oxygen isotopes, and the degree of U-Pb concordance, and to investigate the thermal release characteristics of the xenon. In addition to constraining the terrestrial Pu/U ratio these investigations may allow us to characterise the geochemical behaviour of Pu as a constraint on the earliest crust forming processes.

Formation mechanism of gas bubble superlattice in UMo metal fuels: Phase-field modeling investigation

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

Hu, Shenyang; Burkes, Douglas E.; Lavender, Curt A.

2016-07-08

Nano-gas bubble superlattices are often observed in irradiated UMo nuclear fuels. However, the for- mation mechanism of gas bubble superlattices is not well understood. A number of physical processes may affect the gas bubble nucleation and growth; hence, the morphology of gas bubble microstructures including size and spatial distributions. In this work, a phase-field model integrating a first-passage Monte Carlo method to investigate the formation mechanism of gas bubble superlattices was devel- oped. Six physical processes are taken into account in the model: 1) heterogeneous generation of gas atoms, vacancies, and interstitials informed from atomistic simulations; 2) one-dimensional (1-D) migration of interstitials; 3) irradiation-induced dissolution of gas atoms; 4) recombination between vacancies and interstitials; 5) elastic interaction; and 6) heterogeneous nucleation of gas bubbles. We found that the elastic interaction doesn’t cause the gas bubble alignment, and fast 1-D migration of interstitials alongmore » $$\\langle$$110$$\\rangle$$ directions in the body-centered cubic U matrix causes the gas bubble alignment along $$\\langle$$110$$\\rangle$$ directions. It implies that 1-D interstitial migration along [110] direction should be the primary mechanism of a fcc gas bubble superlattice which is observed in bcc UMo alloys. Simulations also show that fission rates, saturated gas concentration, and elastic interaction all affect the morphology of gas bubble microstructures.« less

An experimental investigation of (UF-235)6 fission nuclear-pumped lasers

NASA Technical Reports Server (NTRS)

Miley, G. H.

1979-01-01

A UF6 handling system was designed for use in conjunction with the existing nuclear-pumped laser vacuum system at a nuclear reactor laboratory to perform the experiments described above. A modification to separate the gas fill system from the vacuum system and thus greatly reduce its volume is described as well as operating procedures for the first controlled nuclear pumping experiments with UF6 vapor contained in the laser cell.

THE ARMOUR DUST FUELED REACTOR (ADFR)

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

Krucoff, D.

1958-01-01

The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

Patent Family Databases.

ERIC Educational Resources Information Center

Simmons, Edlyn S.

1985-01-01

Reports on retrieval of patent information online and includes definition of patent family, basic and equivalent patents, "parents and children" applications, designated states, patent family databases--International Patent Documentation Center, World Patents Index, APIPAT (American Petroleum Institute), CLAIMS (IFI/Plenum). A table…

Multiple tube premixing device

DOEpatents

Uhm, Jong Ho; Naidu, Balachandar; Ziminksy, Willy Steve; Kraemer, Gilbert Otto; Yilmaz, Ertan; Lacy, Benjamin; Stevenson, Christian; Felling, David

2013-08-13

The present application provides a premixer for a combustor. The premixer may include a fuel plenum with a number of fuel tubes and a burner tube with a number of air tubes. The fuel tubes extend about the air tubes.

Multiple tube premixing device

DOEpatents

Uhm, Jong Ho; Varatharajan, Balachandar; Ziminsky, Willy Steve; Kraemer, Gilbert Otto; Yilmaz, Ertan; Lacy, Benjamin; Stevenson, Christian; Felling, David

2012-12-11

The present application provides a premixer for a combustor. The premixer may include a fuel plenum with a number of fuel tubes and a burner tube with a number of air tubes. The fuel tubes extend about the air tubes.

Sorbent-Based Atmosphere Revitalization System

NASA Technical Reports Server (NTRS)

Knox, James C (Inventor); Miller, Lee A. (Inventor)

2017-01-01

The present invention is a sorbent-based atmosphere revitalization (SBAR) system using treatment beds each having a bed housing, primary and secondary moisture adsorbent layers, and a primary carbon dioxide adsorbent layer. Each bed includes a redirecting plenum between moisture adsorbent layers, inlet and outlet ports connected to inlet and outlet valves, respectively, and bypass ports connected to the redirecting plenums. The SBAR system also includes at least one bypass valve connected to the bypass ports. An inlet channel connects inlet valves to an atmosphere source. An outlet channel connects the bypass valve and outlet valves to the atmosphere source. A vacuum channel connects inlet valves, the bypass valve and outlet valves to a vacuum source. In use, one bed treats air from the atmosphere source while another bed undergoes regeneration. During regeneration, the inlet, bypass, and outlet valves sequentially open to the vacuum source, removing accumulated moisture and carbon dioxide.

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

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

Todreas, N.E.; Cheng, S.K.; Basehore, K.

1984-08-01

This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized.more » Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.« less

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1959-12-15

A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

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

Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.

2015-07-01

First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes amore » main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was conducted successfully between 2008 and 2010 on 5 high burn-up MOX fuel rods and 2 very high burn-up UO{sub 2} fuel rods in LECA Facility at Cadarache Centre. An improvement of this sensor has been proposed, allowing us to divide by two the uncertainty on the pressure measurement. In the case of hot-cell measurements, viscous liquid can be used to couple the sensor with the rod. For gas content with a pressure exceeding 15 bars and a 10% Xe/Kr ratio, such coupling may reduce relative acoustic method accuracy by ±7% for pressure measurement result and ±0.25 % for the assessment of gas composition. These results make it possible to demonstrate the feasibility of the technique on LWR fuel rods. The transducer and the associated methodology are now operational for non-destructive measurements in hot lab facilities and allow characterising the fission gas without puncturing the fuel rods. Up to now, any other non-destructive method can be proposed. A next step will be the development of an industrial application in a fuel storage pool in order to perform a large number of measurements on a fuel assembly in nuclear plants.« less

Co-Flow Hollow Cathode Technology

NASA Technical Reports Server (NTRS)

Hofer, Richard R.; Goebel, Dan M.

2011-01-01

Hall thrusters utilize identical hollow cathode technology as ion thrusters, yet must operate at much higher mass flow rates in order to efficiently couple to the bulk plasma discharge. Higher flow rates are necessary in order to provide enough neutral collisions to transport electrons across magnetic fields so that they can reach the discharge. This higher flow rate, however, has potential life-limiting implications for the operation of the cathode. A solution to the problem involves splitting the mass flow into the hollow cathode into two streams, the internal and external flows. The internal flow is fixed and set such that the neutral pressure in the cathode allows for a high utilization of the emitter surface area. The external flow is variable depending on the flow rate through the anode of the Hall thruster, but also has a minimum in order to suppress high-energy ion generation. In the co-flow hollow cathode, the cathode assembly is mounted on thruster centerline, inside the inner magnetic core of the thruster. An annular gas plenum is placed at the base of the cathode and propellant is fed throughout to produce an azimuthally symmetric flow of gas that evenly expands around the cathode keeper. This configuration maximizes propellant utilization and is not subject to erosion processes. External gas feeds have been considered in the past for ion thruster applications, but usually in the context of eliminating high energy ion production. This approach is adapted specifically for the Hall thruster and exploits the geometry of a Hall thruster to feed and focus the external flow without introducing significant new complexity to the thruster design.

Use of CFD to predict trapped gas excitation as source of vibration and noise in screw compressors

NASA Astrophysics Data System (ADS)

Willie, James

2017-08-01

This paper investigates the source of noise in oil free screw compressors mounted on highway trucks and driven by a power take-off (PTO) transmission system. Trapped gas at the discharge side is suggested as possible source of the excitation of low frequency torsional resonance in these compressors that can lead to noise and vibration. Measurements and lumped mass torsional models have shown low frequency torsional resonance in the drive train of these compressors when they are mounted on trucks. This results in high torque peak at the compressor input shaft and in part to pulsating noise inside the machine. The severity of the torque peak depends on the amplitude of the input torque fluctuation from the drive (electric motor or truck engine). This in turn depends on the prop-shaft angle. However, the source of the excitation of this low torsional resonance inside the machine is unknown. Using CFD with mesh motion at every 1° rotation of the rotors, it is shown that the absence of a pressure equalizing chamber at the discharge can lead to trapped gas creation, which can lead to over-compression, over-heating of the rotors, and to high pressure pulsations at the discharge. Over-compression can lead to shock wave generation at the discharge plenum and the pulsation in pressure can lead to noise generation. In addition, if the frequency of the pressure pulsation in the low frequency range coincides with the first torsional frequency of the drive train the first torsional resonance mode can be excited.

RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

NASA Astrophysics Data System (ADS)

Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

Performance Analysis of a Modular Small-Diamter Air Distribution System

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

Poerschke, Andrew; Rudd, Armin

2016-03-01

This report investigates the feasibility of using a home-run manifold small-diameter duct system to provide space conditioning air to individual thermal zones in a low-load home. This compact layout allows duct systems to be brought easily within conditioned space via interior partition walls. Centrally locating the air handler unit in the house significantly reduces duct lengths. The plenum box is designed so that each connected duct receives an equal amount of airflow, regardless of the duct position on the box. Furthermore, within a reasonable set of length restrictions, each duct continues to receive similar airflow. The design method uses anmore » additive approach to reach the total needed zonal airflow. Once the cubic feet per minute needed to satisfy the thermal load of a zone has been determined, the total number of duct runs to a zone can be calculated by dividing the required airflow by the standard airflow from each duct. The additive approach greatly simplifies the design effort and reduces the potential for duct design mistakes to be made. Measured results indicate that this plenum design can satisfy the heating load. However, the total airflow falls short of satisfying the cooling load in a hypothetical building. Static pressure inside the plenum box of 51.5 Pa limited the total airflow of the attached mini-split heat pump blower, thus limiting the total thermal capacity. Fan energy consumption is kept to 0.16 to 0.22 watt/CFM by using short duct runs and smooth duct material.« less

The role of inertial fusion energy in the energy marketplace of the 21st century and beyond

NASA Astrophysics Data System (ADS)

John Perkins, L.

The viability of inertial fusion in the 21st century and beyond will be determined by its ultimate cost, complexity, and development path relative to other competing, long term, primary energy sources. We examine this potential marketplace in terms of projections for population growth, energy demands, competing fuel sources and environmental constraints (CO 2), and show that the two competitors for inertial fusion energy (IFE) in the medium and long term are methane gas hydrates and advanced, breeder fission; both have potential fuel reserves that will last for thousands of years. Relative to other classes of fusion concepts, we argue that the single largest advantage of the inertial route is the perception by future customers that the IFE fusion power core could achieve credible capacity factors, a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. In particular, we show that the size, cost and complexity of the IFE reactor chamber is little different to a fission reactor vessel of the same thermal power. Therefore, relative to fission, because of IFE's tangible advantages in safety, environment, waste disposal, fuel supply and proliferation, our research in advanced targets and innovative drivers can lead to a certain, reduced-size driver at which future utility executives will be indifferent to the choice of an advanced fission plant or an advanced IFE power plant; from this point on, we have a competitive commercial product. Finally, given that the major potential customer for energy in the next century is the present developing world, we put the case for future IFE "reservations" which could be viable propositions providing sufficient reliability and redundancy can be realized for each modular reactor unit.

Progress on Fabrication of Planar Diffusion Couples with Representative TRISO PyC/SiC Microstructure

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

Hunn, John D.; Jolly, Brian C.; Gerczak, Tyler J.

Release of fission products from tristructural-isotropic (TRISO) coated particle fuel limits the fuel’s operational lifetime and creates potential safety and maintenance concerns. A need for diffusion analysis in representative TRISO layers exists to provide fuel performance models with high fidelity data to improve fuel performance and efficiency. An effort has been initiated to better understand fission product transport in, and release from, quality TRISO fuel by investigating diffusion couples with representative pyrocarbon (PyC) and silicon carbide (SiC). Here planar PyC/SiC diffusion couples are being developed with representative PyC/SiC layers using a fluidized bed chemical vapor deposition (FBCVD) system identical tomore » those used to produce laboratory-scale TRISO fuel for the Advanced Gas Reactor Fuel Qualification and Development Program’s (AGR) first fuel irradiation. The diffusivity of silver, the silver and palladium system, europium, and strontium in the PyC/SiC will be studied at elevated temperatures and under high temperature neutron irradiation. The study also includes a comparative study of PyC/SiC diffusion couples with varying TRISO layer properties to understand the influence of SiC microstructure (grain size) and the PyC/SiC interface on fission product transport. The first step in accomplishing these goals is the development of the planar diffusion couples. The diffusion couple construction consists of multiple steps which includes fabrication of the primary PyC/SiC structures with targeted layer properties, introduction of fission product species and seal coating to create an isolated system. Coating development has shown planar PyC/SiC diffusion couples with similar properties to AGR TRISO fuel can be produced. A summary of the coating development process, characterization methods, and status are presented.« less

Thermophysical properties of multi-shock compressed dense argon.

PubMed

Chen, Q F; Zheng, J; Gu, Y J; Chen, Y L; Cai, L C; Shen, Z J

2014-02-21

In contrast to the single shock compression state that can be obtained directly via experimental measurements, the multi-shock compression states, however, have to be calculated with the aid of theoretical models. In order to determine experimentally the multiple shock states, a diagnostic approach with the Doppler pins system (DPS) and the pyrometer was used to probe multiple shocks in dense argon plasmas. Plasma was generated by a shock reverberation technique. The shock was produced using the flyer plate impact accelerated up to ∼6.1 km/s by a two-stage light gas gun and introduced into the plenum argon gas sample, which was pre-compressed from the environmental pressure to about 20 MPa. The time-resolved optical radiation histories were determined using a multi-wavelength channel optical transience radiance pyrometer. Simultaneously, the particle velocity profiles of the LiF window was measured with multi-DPS. The states of multi-shock compression argon plasma were determined from the measured shock velocities combining the particle velocity profiles. We performed the experiments on dense argon plasmas to determine the principal Hugonoit up to 21 GPa, the re-shock pressure up to 73 GPa, and the maximum measure pressure of the fourth shock up to 158 GPa. The results are used to validate the existing self-consistent variational theory model in the partial ionization region and create new theoretical models.

Thermophysical properties of multi-shock compressed dense argon

NASA Astrophysics Data System (ADS)

Chen, Q. F.; Zheng, J.; Gu, Y. J.; Chen, Y. L.; Cai, L. C.; Shen, Z. J.

2014-02-01

In contrast to the single shock compression state that can be obtained directly via experimental measurements, the multi-shock compression states, however, have to be calculated with the aid of theoretical models. In order to determine experimentally the multiple shock states, a diagnostic approach with the Doppler pins system (DPS) and the pyrometer was used to probe multiple shocks in dense argon plasmas. Plasma was generated by a shock reverberation technique. The shock was produced using the flyer plate impact accelerated up to ˜6.1 km/s by a two-stage light gas gun and introduced into the plenum argon gas sample, which was pre-compressed from the environmental pressure to about 20 MPa. The time-resolved optical radiation histories were determined using a multi-wavelength channel optical transience radiance pyrometer. Simultaneously, the particle velocity profiles of the LiF window was measured with multi-DPS. The states of multi-shock compression argon plasma were determined from the measured shock velocities combining the particle velocity profiles. We performed the experiments on dense argon plasmas to determine the principal Hugonoit up to 21 GPa, the re-shock pressure up to 73 GPa, and the maximum measure pressure of the fourth shock up to 158 GPa. The results are used to validate the existing self-consistent variational theory model in the partial ionization region and create new theoretical models.

Superheavy element chemistry at GARIS

NASA Astrophysics Data System (ADS)

Haba, Hiromitsu

2016-12-01

A gas-jet transport system has been installed to the RIKEN GAs-filled Recoil Ion Separator, GARIS to start up SuperHeavy Element (SHE) chemistry. This system is a promising approach for exploring new frontiers in SHE chemistry: background radioactivities from unwanted by-products are suppressed, a high gas-jet transport yield is achieved, and new chemical reactions can be investigated. Useful radioisotopes of 261Rfa,b, 262Db, and 265Sga,b for chemical studies were produced in the reactions of 248Cm(18O,5n)261Rfa,b, 248Cm(19F,5n)262Db, and 248Cm(22Ne,5n)265Sga,b, respectively. They were successfully extracted to a chemistry laboratory by the gas-jet method. Production and decay properties of 261Rfa,b, 262Db, and 265Sga,b were investigated in detail with the rotating wheel apparatus for α- and spontaneous fission spectrometry. Present status and perspectives of the SHE chemistry at GARIS are also briefly presented.

NEUTRONIC REACTOR CORE

DOEpatents

Thomson, W.B.; Corbin, A. Jr.

1961-07-18

An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

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

Greenwood, L. R.; Cantaloub, M. G.; Burnett, J. L.

PNNL has developed two low-background gamma-ray spectrometers in a new shallow underground laboratory, thereby significantly improving its ability to detect low levels of gamma-ray emitting fission or activation products in airborne particulate in samples from the IMS (International Monitoring System). Furthermore, the combination of cosmic veto panels, dry nitrogen gas to reduce radon and low background shielding results in a reduction of the background count rate by about a factor of 100 compared to detectors operating above ground at our laboratory.

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

Loyalka, Sudarshan

High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuelmore » was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.« less

A Summary of Closed Brayton Cycle Development Activities at NASA

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2009-01-01

NASA has been involved in the development of Closed Brayton Cycle (CBC) power conversion technology since the 1960's. CBC systems can be coupled to reactor, isotope, or solar heat sources and offer the potential for high efficiency, long life, and scalability to high power. In the 1960's and 1970's, NASA and industry developed the 10 kW Brayton Rotating Unit (BRU) and the 2 kW mini-BRU demonstrating technical feasibility and performance, In the 1980's, a 25 kW CBC Solar Dynamic (SD) power system option was developed for Space Station Freedom and the technology was demonstrated in the 1990's as part of the 2 kW SO Ground Test Demonstration (GTD). Since the early 2000's, NASA has been pursuing CBC technology for space reactor applications. Before it was cancelled, the Jupiter Icy Moons Orbiter (HMO) mission was considering a 100 kWclass CBC system coupled to a gas-cooled fission reactor. Currently, CBC technology is being explored for Fission Surface Power (FSP) systems to provide base power on the moon and Mars. These recent activities have resulted in several CBC-related technology development projects including a 50 kW Alternator Test Unit, a 20 kW Dual Brayton Test Loop, a 2 kW Direct Drive Gas Brayton Test Loop, and a 12 kW FSP Power Conversion Unit design.

Enhanced trigger for the NIFFTE fissionTPC in presence of high-rate alpha backgrounds

NASA Astrophysics Data System (ADS)

Bundgaard, Jeremy; Niffte Collaboration

2015-10-01

Nuclear physics and nuclear energy communities call for new, high precision measurements to improve existing fission models and design next generation reactors. The Neutron Induced Fission Fragment Tracking experiment (NIFFTE) has developed the fission Time Projection Chamber (fissionTPC) to measure neutron induced fission with unrivaled precision. The fissionTPC is annually deployed to the Weapons Neutron Research facility at Los Alamos Neutron Science Center where it operates with a neutron beam passing axially through the drift volume, irradiating heavy actinide targets to induce fission. The fissionTPC was developed at the Lawrence Livermore National Laboratory's TPC lab, where it measures spontaneous fission from radioactive sources to characterize detector response, improve performance, and evolve the design. To measure 244Cm, we've developed a fission trigger to reduce the data rate from alpha tracks while maintaining a high fission detection efficiency. In beam, alphas from 239Pu are a large background when detecting fission fragments; implementing the fission trigger will greatly reduce this background. The implementation of the cathode fission trigger in the fissionTPC will be presented along with a detailed study of its efficiency.

A physical description of fission product behavior fuels for advanced power reactors.

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

Kaganas, G.; Rest, J.; Nuclear Engineering Division

2007-10-18

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuelsmore » under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.« less

A Conspectus on US Energy

NASA Astrophysics Data System (ADS)

Hayden, Howard

2009-05-01

Until about 1850, the energy used in the US came almost exclusively from firewood. Now we use petroleum, coal, natural gas, nuclear fission, indirect solar energy (biomass, hydro, and wind), geothermal energy, and direct solar energy (solar/thermal, solar/thermal/electric, and photovoltaics). Compared to our ancestors in 1850, we use over 40 times as much energy, of which only about 6 percent is from solar sources, versus 100% in 1850. On a per-capita basis we use about 3.1 times as much energy, in spite of the modern conveniences that to Abraham Lincoln would seem unthinkably lavish. The US uses about 107 EJ of primary energy annually, equivalent to 3.4 TW around-the-clock average power. About 40 percent of that energy goes toward production of electricity. Approximately 2 EJ of heat is obtained from combined heat-and-power plants that produce about 10^9 kWh (3.6 PJ) of electricity. (N.B.: hydro and wind do not involve heat-to-work conversion. By custom, the electrical energy produced by wind and hydro is multiplied by about 3 to generate an as-if quantity of primary energy.) When account is taken of how the electricity is distributed, industry uses 33 percent of the primary energy, followed by transportation (28%), residences (21%), and commercial establishments (18%). ``Fossil fuels'' (coal, oil, and natural gas) account for about 85 percent of our primary energy. Nuclear energy accounts for about 8%. Biomass and hydro, the venerable solar-derived sources, account for about 7%. Geothermal, wind, and direct solar energy account for about 0.4%. This talk will discuss prospects for various alternative sources, including nuclear fission and T. Boone Pickens' plan to displace imported petroleum indirectly by substituting wind for natural gas.

Cooling molten salt reactors using "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

2014-08-01

This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

Classical Molecular Dynamics Simulation of Nuclear Fuel

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

Devanathan, Ram; Krack, Matthias; Bertolus, Marjorie

2015-10-10

Molecular dynamics simulation is well suited to study primary damage production by irradiation, defect interactions with fission gas atoms, gas bubble nucleation, grain boundary effects on defect and gas bubble evolution in nuclear fuel, and the resulting changes in thermo-mechanical properties. In these simulations, the forces on the ions are dictated by interaction potentials generated by fitting properties of interest to experimental data. The results obtained from the present generation of potentials are qualitatively similar, but quantitatively different. There is a need to refine existing potentials to provide a better representation of the performance of polycrystalline fuel under a varietymore » of operating conditions, and to develop models that are equipped to handle deviations from stoichiometry. In addition to providing insights into fundamental mechanisms governing the behaviour of nuclear fuel, MD simulations can also provide parameters that can be used as inputs for mesoscale models.« less

Superconducting RF Linacs Driving Subcritical Reactors for Profitable Disposition of Surplus Weapons-grade Plutonium

NASA Astrophysics Data System (ADS)

Cummings, Mary Anne; Johnson, Rolland

Acceptable capital and operating costs of high-power proton accelerators suitable for profitable commercial electric-power and process-heat applications have been demonstrated. However, studies have pointed out that even a few hundred trips of an accelerator lasting a few seconds would lead to unacceptable thermal stresses as each trip causes fission to be turned off in solid fuel structures found in conventional reactors. The newest designs based on the GEM*STAR concept take such trips in stride by using molten-salt fuel, where fuel pin fatigue is not an issue. Other aspects of the GEM*STAR concept which address all historical reactor failures include an internal spallation neutron target and high temperature molten salt fuel with continuous purging of volatile radioactive fission products such that the reactor contains less than a critical mass and almost a million times fewer volatile radioactive fission products than conventional reactors. GEM*STAR is a reactor that without redesign will burn spent nuclear fuel, natural uranium, thorium, or surplus weapons material. It will operate without the need for a critical core, fuel enrichment, or reprocessing making it an excellent candidate for export. As a first application, the design for a pilot plant is described for the profitable disposition of surplus weapons-grade plutonium by using process heat to produce green diesel fuel for the Department of Defense (DOD) from natural gas and renewable carbon.

High-spin structure of 134Xe

NASA Astrophysics Data System (ADS)

Vogt, A.; Birkenbach, B.; Reiter, P.; Blazhev, A.; Siciliano, M.; Valiente-Dobón, J. J.; Wheldon, C.; Bazzacco, D.; Bowry, M.; Bracco, A.; Bruyneel, B.; Chakrawarthy, R. S.; Chapman, R.; Cline, D.; Corradi, L.; Crespi, F. C. L.; Cromaz, M.; de Angelis, G.; Eberth, J.; Fallon, P.; Farnea, E.; Fioretto, E.; Freeman, S. J.; Gadea, A.; Geibel, K.; Gelletly, W.; Gengelbach, A.; Giaz, A.; Görgen, A.; Gottardo, A.; Hayes, A. B.; Hess, H.; Hua, H.; John, P. R.; Jolie, J.; Jungclaus, A.; Korten, W.; Lee, I. Y.; Leoni, S.; Liang, X.; Lunardi, S.; Macchiavelli, A. O.; Menegazzo, R.; Mengoni, D.; Michelagnoli, C.; Mijatović, T.; Montagnoli, G.; Montanari, D.; Napoli, D.; Pearson, C. J.; Pellegri, L.; Podolyák, Zs.; Pollarolo, G.; Pullia, A.; Radeck, F.; Recchia, F.; Regan, P. H.; Şahin, E.; Scarlassara, F.; Sletten, G.; Smith, J. F.; Söderström, P.-A.; Stefanini, A. M.; Steinbach, T.; Stezowski, O.; Szilner, S.; Szpak, B.; Teng, R.; Ur, C.; Vandone, V.; Ward, D.; Warner, D. D.; Wiens, A.; Wu, C. Y.

2016-05-01

Detailed spectroscopic information on the N ˜82 nuclei is necessary to benchmark shell-model calculations in the region. The nuclear structure above long-lived isomers in 134Xe is investigated after multinucleon transfer (MNT) and actinide fission. Xenon-134 was populated as (i) a transfer product in 238U+ 136Xe and 208Pb+ 136Xe MNT reactions and (ii) as a fission product in the 238U+ 136Xe reaction employing the high-resolution Advanced Gamma Tracking Array (AGATA). Trajectory reconstruction has been applied for the complete identification of beamlike transfer products with the magnetic spectrometer PRISMA. The 198Pt 136Xe MNT reaction was studied with the γ -ray spectrometer GAMMASPHERE in combination with the gas detector array Compact Heavy Ion Counter (CHICO). Several high-spin states in 134Xe on top of the two long-lived isomers are discovered based on γ γ -coincidence relationships and information on the γ -ray angular distributions as well as excitation energies from the total kinetic energy loss and fission fragments. The revised level scheme of 134Xe is extended up to an excitation energy of 5.832 MeV with tentative spin-parity assignments up to 16+. Previous assignments of states above the 7- isomer are revised. Latest shell-model calculations employing two different effective interactions reproduce the experimental findings and support the new spin and parity assignments.

Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2005-01-01

The EFF-TF provides a facility to experimentally evaluate thermal hydraulic issues through the use of highly effective non-nuclear testing. These techniques provide a rapid, more cost effective method of evaluating designs and support development risk mitigation when concerns are associated with non-nuclear aspects of space nuclear systems. For many systems, electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004. Initial evaluation of the SAFE-100a (19 module stainless steel/sodium heat pipe reactor with integral gas neat exchanger) was performed with tests up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium SAFE-100 heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37-fuel pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to field a near term space nuclear system. Efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

Exhaust gas treatment in testing nuclear rocket engines

NASA Astrophysics Data System (ADS)

Zweig, Herbert R.; Fischler, Stanley; Wagner, William R.

1993-01-01

With the exception of the last test series of the Rover program, Nuclear Furnace 1, test-reactor and rocket engine hydrogen gas exhaust generated during the Rover/NERVA program was released directly to the atmosphere, without removal of the associated fission products and other radioactive debris. Current rules for nuclear facilities (DOE Order 5480.6) are far more protective of the general environment; even with the remoteness of the Nevada Test Site, introduction of potentially hazardous quantities of radioactive waste into the atmosphere must be scrupulously avoided. The Rocketdyne treatment concept features a diffuser to provide altitude simulation and pressure recovery, a series of heat exchangers to gradually cool the exhaust gas stream to 100 K, and an activated charcoal bed for adsorption of inert gases. A hydrogen-gas fed ejector provides auxiliary pumping for startup and shutdown of the engine. Supplemental filtration to remove particulates and condensed phases may be added at appropriate locations in the system. The clean hydrogen may be exhausted to the atmosphere and flared, or the gas may be condensed and stored for reuse in testing. The latter approach totally isolates the working gas from the environment.

A molten Salt Am242M Production Reactor for Space Applications

NASA Technical Reports Server (NTRS)

Emrich, William

2005-01-01

The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

TECHNICAL SCOPE OF GAS-COOLED REACTOR FUEL ELEMENT IRRADIATION PROGRAM

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

None

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

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

Scofield, C.M.; Des Champs, N.H.

This article examines a design concept for classroom air conditioning systems that guarantees minimum ventilation rates are met. The topics of the article include new ventilation requirements, design concept, outside air induction diffuser, low-velocity ducts and plenums, the relationship of humidity to school absenteeism rates, retrofit applications, and saving energy.

Cooling Computers.

ERIC Educational Resources Information Center

Birken, Marvin N.

1967-01-01

Numerous decisions must be made in the design of computer air conditioning, each determined by a combination of economics, physical, and esthetic characteristics, and computer requirements. Several computer air conditioning systems are analyzed--(1) underfloor supply and overhead return, (2) underfloor plenum and overhead supply with computer unit…

Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.

2007-01-01

In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

Extinct Plutonium Geochemistry of Ancient Hadean Zircons

NASA Astrophysics Data System (ADS)

Turner, G.; Gilmour, J.; Crowther, S.; Busfield, A.; Mojzsis, S.; Harrison, M.

2005-12-01

The abundance of 244Pu in the early solar system has important implications for r-process nucleosynthesis and models of noble gas transport within the Earth's mantle. Our recent discovery(1) of xenon isotopes from the in-situ decay of 244Pu in ancient Jack Hills zircons promises to provide a new time-sensitive window on the first 500 Ma of Earth history. We have extended this initial work by the use of resonance ioniisation mass spectrometry to analyse xenon released by stepped heating from 17 individual zircons with Pb-Pb ages in the range 3.95 to 4.18 Ga. Our immediate objectives are to determine the causes of variations in the inferred Pu/U ratios and in the longer term to determine the initial Pu/U ratio of the Earth. The Pu/U ratios calculated for individual zircons may be expected to vary as a result of igneous fractionation and also from differential loss of Pu and U fission xenon in the last 4 Ga. We have studied the effects of xenon loss by irradiating the zircons with thermal neutrons to generate xenon from 235U neutron fission in order to determine U/Xe ratios and apparent ages. 131Xe/134Xe and 132Xe/134Xe ratios can be used to calculate the relative contributions from 244Pu and 238U spontaneous fission and 235U neutron fission. The measured Pu/U ratios (back calculated to 4.56 Ga on the basis of the individual Pb-Pb ages) range from zero to 0.012. The highest ratio in our initial study was 0.008 (note that the published ratio has been revised upwards on the basis of improved decay parameters for 238U spontaneous fission). Comparison of Pb-Pb and U-Xe ages indicate varying amounts of xenon loss, over 50% in some cases. While this accounts for some of the variability in the inferred Pu/U, igneous fractionation may also play a part, and we are currently attempting to investigate this by a comparison with REE abundances. Reference: (1) Turner et al. (2004) Science, 306, 89-91.

KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

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

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key resultsmore » from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program is also discussed.« less

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

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

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on keymore » results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed.« less

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

DOE PAGES

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; ...

2017-09-10

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on keymore » results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed.« less

Extreme population inversion in the fragments formed by UV photoinduced S-H bond fission in 2-thiophenethiol.

PubMed

Ingle, Rebecca A; Karsili, Tolga N V; Dennis, Gregg J; Staniforth, Michael; Stavros, Vasilios G; Ashfold, Michael N R

2016-04-28

H atom loss following near ultraviolet photoexcitation of gas phase 2-thiophenethiol molecules has been studied experimentally, by photofragment translational spectroscopy (PTS) methods, and computationally, by ab initio electronic structure calculations. The long wavelength (277.5 ≥ λ(phot) ≥ 240 nm) PTS data are consistent with S-H bond fission after population of the first (1)πσ* state. The partner thiophenethiyl (R) radicals are formed predominantly in their first excited Ã(2)A' state, but assignment of a weak signal attributable to H + R(X˜(2)A'') products allows determination of the S-H bond strength, D0 = 27,800 ± 100 cm(-1) and the Ã-X˜ state splitting in the thiophenethiyl radical (ΔE = 3580 ± 100 cm(-1)). The deduced population inversion between the à and X˜ states of the radical reflects the non-planar ground state geometry (wherein the S-H bond is directed near orthogonal to the ring plane) which, post-photoexcitation, is unable to planarise sufficiently prior to bond fission. This dictates that the dissociating molecules follow the adiabatic fragmentation pathway to electronically excited radical products. π* ← π absorption dominates at shorter excitation wavelengths. Coupling to the same (1)πσ* potential energy surface (PES) remains the dominant dissociation route, but a minor yield of H atoms attributable to a rival fragmentation pathway is identified. These products are deduced to arise via unimolecular decay following internal conversion to the ground (S0) state PES via a conical intersection accessed by intra-ring C-S bond extension. The measured translational energy disposal shows a more striking change once λ(phot) ≤ 220 nm. Once again, however, the dominant decay pathway is deduced to be S-H bond fission following coupling to the (1)πσ* PES but, in this case, many of the evolving molecules are deduced to have sufficiently near-planar geometries to allow passage through the conical intersection at extended S-H bond lengths and dissociation to ground (X˜) state radical products. The present data provide no definitive evidence that complete ring opening can compete with fast S-H bond fission following near UV photoexcitation of 2-thiophenethiol.

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

Duct attachment and extension for an air conditioning unit

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

Lang, R.D.; Frenia, F.J.

1986-12-16

An apparatus is described for attaching a fixed duct extension to the discharge opening of an air conditioning unit, the unit slidably inserted in and removed from a fixed through-the-wall sleeve, for supplying conditioned air to the space containing the unit and an adjacent space comprising: a discharge plenum assembly adapted to be connected to the unit encase the discharge opening. The discharge plenum assembly defines an air flow path for the conditioned air discharged from the unit and includes a first housing member having a forward wall, a rear wall, and a pair of opposed side walls joining themore » front wall to the rear wall, and a second housing member having a top wall connected to a front wall. The top wall and the front wall are fixedly attached to the rear wall and the forward wall respectively of the first housing member and forming a duct outlet in one of the side walls. The top wall and the front wall of the second housing member and one of the pair of opposed side walls of the lower housing member having longitudinal flanges extending therefrom forming a C-like flange; a bracket removably secured to the through the wall sleeve having an outwardly extending flange member at the top of the bracket; and a duct extension means secured to the outwardly extending flange of the bracket near one end and to the wall of the adjacent space at the opposite end. The duct extension means has a collar at one end configured to engage with the C-like flange whereby the unit with the discharge plenum assembly attached thereto slidably engages with and disengages from the through-the-wall sleeve while the duct extension is secured to the bracket.« less

Spectral properties of gaseous uranium hexafluoride at high temperature

NASA Technical Reports Server (NTRS)

Krascella, N. L.

1980-01-01

A study to determine relative spectral emission and spectral absorption data for UF6-argon mixtures at elevated temperatures is discussed. These spectral data are required to assist in the theoretical analysis of radiation transport in the nuclear fuel-buffer gas region of a plasma core reactor. Relative emission measurements were made for UF6-argon mixtures over a range of temperatures from 650 to 1900 K and in the wavelength range from 600 to 5000 nanometers. All emission results were determined for a total pressure of 1.0 atm. Uranium hexafluoride partial pressures varied from about 3.5 to 12.7 mm Hg. Absorption measurements were attempted at 600, 625, 650 and 675 nanometers for a temperature of 1000 K. The uranium partial pressure for these determinations was 25 mm Hg. The results exhibit appreciable emission for hot UF6-argon mixtures at wavelengths between 600 and 1800 nanometers and no measurable absorption. The equipment used to evaluate the spectral properties of the UF6-argon mixtures included a plasma torch-optical plenum assembly, the monochromator, and the UF6 transfer system. Each is described.

Uniform-burning matrix burner

DOEpatents

Bohn, Mark S.; Anselmo, Mark

2001-01-01

Computer simulation was used in the development of an inward-burning, radial matrix gas burner and heat pipe heat exchanger. The burner and exchanger can be used to heat a Stirling engine on cloudy days when a solar dish, the normal source of heat, cannot be used. Geometrical requirements of the application forced the use of the inward burning approach, which presents difficulty in achieving a good flow distribution and air/fuel mixing. The present invention solved the problem by providing a plenum with just the right properties, which include good flow distribution and good air/fuel mixing with minimum residence time. CFD simulations were also used to help design the primary heat exchanger needed for this application which includes a plurality of pins emanating from the heat pipe. The system uses multiple inlet ports, an extended distance from the fuel inlet to the burner matrix, flow divider vanes, and a ring-shaped, porous grid to obtain a high-temperature uniform-heat radial burner. Ideal applications include dish/Stirling engines, steam reforming of hydrocarbons, glass working, and any process requiring high temperature heating of the outside surface of a cylindrical surface.

Low-background gamma-ray spectrometry for the international monitoring system

DOE PAGES

Greenwood, L. R.; Cantaloub, M. G.; Burnett, J. L.; ...

2016-12-28

PNNL has developed two low-background gamma-ray spectrometers in a new shallow underground laboratory, thereby significantly improving its ability to detect low levels of gamma-ray emitting fission or activation products in airborne particulate in samples from the IMS (International Monitoring System). Furthermore, the combination of cosmic veto panels, dry nitrogen gas to reduce radon and low background shielding results in a reduction of the background count rate by about a factor of 100 compared to detectors operating above ground at our laboratory.