NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

DOEpatents

McCorkle, W.H.

1960-02-23

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

Coolant mass flow equalizer for nuclear fuel

DOEpatents

Betten, Paul R.

1978-01-01

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

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

DOEpatents

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

1959-01-27

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

FUEL ELEMENT

DOEpatents

Bean, R.W.

1963-11-19

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

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

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

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

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

FUEL ELEMENT FOR NEUTRONIC REACTORS

DOEpatents

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

1961-01-17

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

Emergency cooling analysis for the loss of coolant malfunction

NASA Technical Reports Server (NTRS)

Peoples, J. A.

1972-01-01

This report examines the dynamic response of a conceptual space power fast-spectrum lithium cooled reactor to the loss of coolant malfunction and several emergency cooling concepts. The results show that, following the loss of primary coolant, the peak temperatures of the center most 73 fuel elements can range from 2556 K to the region of the fuel melting point of 3122 K within 3600 seconds after the start of the accident. Two types of emergency aftercooling concepts were examined: (1) full core open loop cooling and (2) partial core closed loop cooling. The full core open loop concept is a one pass method of supplying lithium to the 247 fuel pins. This method can maintain fuel temperature below the 1611 K transient damage limit but requires a sizable 22,680-kilogram auxiliary lithium supply. The second concept utilizes a redundant internal closed loop to supply lithium to only the central area of each hexagonal fuel array. By using this method and supplying lithium to only the triflute region, fuel temperatures can be held well below the transient damage limit.

NUCLEAR REACTOR

DOEpatents

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

1958-12-01

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

SELECTIVE SEPARATION OF URANIUM FROM FERRITIC STAINLESS STEELS

DOEpatents

Beaver, R.J.; Cherubini, J.H.

1963-05-14

A process is described for separating uranium from a nuclear fuel element comprising a uranium-containing core and a ferritic stainless steel clad by heating said element in a non-carburizing atmosphere at a temperature in the range 850-1050 un. Concent 85% C, rapidly cooling the heated element through the temperature range 815 un. Concent 85% to 650 EC to avoid annealing said steel, and then contacting the cooled element with an aqueous solution of nitric acid to selectively dissolve the uranium. (AEC)

Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

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

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

2009-03-01

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

NEUTRONIC REACTOR

DOEpatents

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

1958-09-01

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

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

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

Thermal modeling in an engine cooling system to control coolant flow for fuel consumption improvement

NASA Astrophysics Data System (ADS)

Park, Sangki; Woo, Seungchul; Kim, Minho; Lee, Kihyung

2017-04-01

The design and evaluation of engine cooling and lubrication systems is generally based on real vehicle tests. Our goal here was to establish an engine heat balance model based on mathematical and interpretive analysis of each element of a passenger diesel engine cooling system using a 1-D numerical model. The purpose of this model is to determine ways of optimizing the cooling and lubrication components of an engine and then to apply these methods to actual cooling and lubrication systems of engines that will be developed in the future. Our model was operated under the New European Driving Cycle (NEDC) mode conditions, which represent the fuel economy evaluation mode in Europe. The flow rate of the cooling system was controlled using a control valve. Our results showed that the fuel efficiency was improved by as much as 1.23 %, cooling loss by 1.35 %, and friction loss by 2.21 % throughout NEDC modes by modification of control conditions.

JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

DOEpatents

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

1959-05-12

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

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)

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

Numerical simulation of a mini PEMFC stack

NASA Astrophysics Data System (ADS)

Liu, Zhixiang; Mao, Zongqiang; Wang, Cheng; Zhuge, Weilin; Zhang, Yangjun

Fuel cell modeling and simulation has aroused much attention recently because it can probe transport and reaction mechanism. In this paper, a computational fuel cell dynamics (CFCD) method was applied to simulate a proton exchange membrane fuel cell (PEMFC) stack for the first time. The air cooling mini fuel cell stack consisted of six cells, in which the active area was 8 cm 2 (2 cm × 4 cm). With reasonable simplification, the computational elements were effectively reduced and allowed a simulation which could be conducted on a personal computer without large-scale parallel computation. The results indicated that the temperature gradient inside the fuel cell stack was determined by the flow rate of the cooling air. If the air flow rate is too low, the stack could not be effectively cooled and the temperature will rise to a range that might cause unstable stack operation.

Secondary fuel delivery system

DOEpatents

Parker, David M.; Cai, Weidong; Garan, Daniel W.; Harris, Arthur J.

2010-02-23

A secondary fuel delivery system for delivering a secondary stream of fuel and/or diluent to a secondary combustion zone located in the transition piece of a combustion engine, downstream of the engine primary combustion region is disclosed. The system includes a manifold formed integral to, and surrounding a portion of, the transition piece, a manifold inlet port, and a collection of injection nozzles. A flowsleeve augments fuel/diluent flow velocity and improves the system cooling effectiveness. Passive cooling elements, including effusion cooling holes located within the transition boundary and thermal-stress-dissipating gaps that resist thermal stress accumulation, provide supplemental heat dissipation in key areas. The system delivers a secondary fuel/diluent mixture to a secondary combustion zone located along the length of the transition piece, while reducing the impact of elevated vibration levels found within the transition piece and avoiding the heat dissipation difficulties often associated with traditional vibration reduction methods.

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

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

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

Design, fabrication, and testing of an external fuel (UO2), full-length thermionic converter

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

The development of a full-length external-fuel thermionic converter for in-pile testing is described. The development program includes out-of-pile performance testing of the fully fueled-converter, using RF-induction heating, before its installation in the in-pile test capsule. The external-fuel converter is cylindrical in shape, and consists of an inner, centrally cooled collector, and an outer emitter surrounded by nuclear fuel. The term full-length denotes that the converter is long enough to extend over the full height of the reactor core. Thus, the converter is not a scaled-down test device, but a full-scale fuel element of the thermionic reactor. The external-fuel converter concept permits a number of different design options, particularly with respect to the fuel composition and shape, and the collector cooling arrangement. The converter described was developed for the Jet Propulsion Laboratory, and is based on their concept for a thermionic reactor with uninsulated collector cooling as previously described. The converter is double-ended, with through-flow cooling, and with ceramic seals and emitter and collector power take-offs at both ends. The design uses a revolver-shaped tungsten emitter body, with the central emitter hole surrounded by six peripheral fuel holes loaded with cylindrical UO2 pellets.

NEUTRONIC REACTOR CONSTRUCTION

DOEpatents

Vernon, H.C.; Goett, J.J.

1958-09-01

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

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.

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

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

McCulloch, R.W.; Post, D.W.; Lovell, R.T.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

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

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

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

2013-07-01

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

SODIUM DEUTERIUM REACTOR

DOEpatents

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

1963-02-26

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

Methods for making a porous nuclear fuel element

DOEpatents

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

2014-12-30

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

Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor

NASA Astrophysics Data System (ADS)

Puji Hastuti, Endiah; Widodo, Surip; Darwis Isnaini, M.; Geni Rina, S.; Syaiful, B.

2018-02-01

TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

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

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

Bauer, Thomas L.; Krause, Michael G

1986-07-01

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

SHAPED FISSIONABLE METAL BODIES

DOEpatents

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

1958-10-14

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

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

DOEpatents

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

1963-12-31

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

A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

NASA Astrophysics Data System (ADS)

Navarro, Jorge

The goal of this study presented is to determine the best available nondestructive technique necessary to collect validation data as well as to determine burnup and cooling time of the fuel elements on-site at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal, the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements nondestructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed were used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results, it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however, in order to enhance the quality of the spectra collected using this scintillator, a deconvolution method was developed. Following the development of the deconvolution method for ATR applications, the technique was tested using one-isotope, multi-isotope, and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr 3 detector in an above the water configuration and deconvolution algorithms.

GEH-4-42, 47; Hot pressed, I and E cooled fuel element irradiation

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

Neidner, R.

1959-11-02

In our continual effort to improve the present fuel elements which are irradiated in the numerous Hanford reactors, we have made what we believe to be a significant improvement in the hot pressing process for jacketing uranium fuel slugs. We are proposing a large scale evaluation testing program in the Hanford reactors but need the vital and basic information on the operating characteristics of this type slug under known and controlled operating conditions. We, therefore, have prepared two typical fuel slugs and will want them irradiated to about 1000 MWD/T exposure (this will require about four to five total cycles).

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

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

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

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

Determination of the steam volume fraction in the event of loss of cooling of the spent fuel storage pool

NASA Astrophysics Data System (ADS)

Sledkov, R. M.; Galkin, I. Yu.; Stepanov, O. E.; Strebnev, N. A.

2017-01-01

When one solves engineering problems related to the cooling of fuel assemblies (FAs) in a spent fuel storage pool (SFSP) and the assessment of nuclear safety of FA storage in an SFSP in the initial event of loss of SFSP cooling, it is essential to determine the coolant density and, consequently, steam volume fractions φ in bundles of fuel elements at a pressure of 0.1-0.5 MPa. Such formulas for calculating φ that remain valid in a wide range of operating parameters and geometric shapes of channels and take the conditions of loss of SFSP cooling into account are currently almost lacking. The results of systematization and analysis of the available formulas for φ are reported in the present study. The calculated values were compared with the experimental data obtained in the process of simulating the conditions of FA cooling in an SFSP in the event of loss of its cooling. Six formulas for calculating the steam volume fraction, which were used in this comparison, were chosen from a total of 11 considered relations. As a result, the formulas producing the most accurate values of φ in the conditions of loss of SFSP cooling were selected. In addition, a relation that allows one to perform more accurate calculations of steam volume fractions in the conditions of loss of SFSP cooling was derived based on the Fedorov formula in the two-group approximation.

AIR COOLED NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Szilard, L.

1958-05-27

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Neutronic fuel element fabrication

DOEpatents

Korton, George

2004-02-24

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

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

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

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

2007-12-01

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

DISCHARGE DEVICE FOR RADIOACTIVE MATERIAL

DOEpatents

Ohlinger, L.A.

1958-09-23

A device is described fur unloading bodies of fissionable material from a neutronic reactor. It is comprised essentially of a wheeled flat car having a receptacle therein containing a liquid coolant fur receiving and cooling the fuel elements as they are discharged from the reactor, and a reciprocating plunger fur supporting the fuel element during discharge thereof prior to its being dropped into the coolant. The flat car is adapted to travel along the face of the reactor adjacent the discharge ends of the coolant tubes.

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

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

Talley, Darren G.

2017-04-01

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

A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

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

Navarro, Jorge

2013-12-01

The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent tomore » the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.« less

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

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

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

The Potential of Different Concepts of Fast Breeder Reactor for the French Fleet Renewal

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

Massara, Simone; Tetart, Philippe; Lecarpentier, David

2006-07-01

The performances of different concepts of Fast Breeder Reactor (Na-cooled, He-cooled and Pb-cooled FBR) for the current French fleet renewal are analyzed in the framework of a transition scenario to a 100% FBR fleet at the end of the 21. century. Firstly, the modeling of these three FBR types by means of a semi-analytical approach in TIRELIRE - STRATEGIE, the EDF fuel cycle simulation code, is presented, together with some validation elements against ERANOS, the French reference code system for neutronic FBR analysis (CEA). Afterwards, performances comparisons are made in terms of maximum deployable power, natural uranium consumption and wastemore » production. The results show that the FBR maximum deployable capacity, independently from the FBR technology, is highly sensitive to the fuel cycle options, like the spent nuclear fuel cooling time or the Minor Actinides management strategy. Thus, some of the key parameters defining the dynamic of FBR deployment are highlighted, to inform the orientation of R and D in the development and optimization of these systems. (authors)« less

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

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

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Dismantlement of the TSF-SNAP Reactor Assembly

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

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

Metallic impurities-silicon carbide interaction in HTGR fuel particles

NASA Astrophysics Data System (ADS)

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

1990-12-01

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

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

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

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

2016-09-01

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

Core cooling under accident conditions at the high-flux beam reactor

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

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« less

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

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

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

Toxicity of irradiated advanced heavy water reactor fuels.

PubMed

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

2013-02-01

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

Process for massively hydriding zirconium--uranium fuel elements

DOEpatents

Katz, N.H.

1973-12-01

A method is described of hydriding uranium-zirconium alloy by heating the alloy in a vacuum, introducing hydrogen and maintaining an elevated temperature until occurrence of the beta--delta phase transformation and isobarically cooling the composition. (Official Gazette)

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William

2013-01-01

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

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

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

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

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

A feasibility study of reactor-based deep-burn concepts.

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

Kim, T. K.; Taiwo, T. A.; Hill, R. N.

2005-09-16

A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team.more » The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling or by operating with lower material temperatures (colder core). Results also showed that the transmutation performance of the one-pass deep-burn concept is sensitive to the initial TRU vector, primarily because longer cooling time reduces the fissile content (Pu-241 specifically.) With a cooling time of 5 years, the TRU consumption increases to 67%, while conversely, with 20-year cooling the TRU consumption is about 58%. For the two-pass DB-MHR (TRU recycling option), a fuel packing fraction of about 30% is required in the second pass (the recycled TRU). It was found that using a heterogeneous core (homogeneous fuel element) concept, the TRU consumption is dependent on the cooling interval before the 2nd pass, again due to Pu-241 decay during the time lag between the first pass fuel discharge and the second pass fuel charge. With a cooling interval of 7 years (5 and 2 years before and after reprocessing) a TRU consumption of 55% is obtained. With an assumed ''no cooling'' interval, the TRU consumption is 63%. By using a cylindrical core to reduce neutron leakage, TRU consumption of the case with 7-year cooling interval increases to 58%. For a two-pass concept using a heterogeneous fuel element (and homogeneous core) with first and second pass volume ratio of 2:1, the TRU consumption is 62.4%. Finally, the repository loading benefits arising from the deep-burn and Inert Matrix Fuel (IMF) concepts were estimated and compared, for the same initial TRU vector. The DB-MHR concept resulted in slightly higher TRU consumption and repository loading benefit compared to the IMF concept (58.1% versus 55.1% for TRU consumption and 2.0 versus 1.6 for estimated repository loading benefit).« less

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.

NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

DOEpatents

Wigner, E.P.; Ohlinger, L.A.

1958-11-18

Reactors of the heterogeneous, graphite moderated, fluid cooled type and shielding and spacing plugs for the coolant channels thereof are reported. In this design, the coolant passages extend horizontally through the moderator structure, accommodating the fuel elements in abutting end-to-end relationship, and have access openings through the outer shield at one face of the reactor to facilitate loading of the fuel elements. In the outer ends of the channels which extend through the shields are provided spacers and shielding plugs designed to offer minimal reslstance to coolant fluid flow while preventing emanation of harmful radiation through the access openings when closed between loadings.

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

NASA Astrophysics Data System (ADS)

Govers, K.; Verwerft, M.

2016-09-01

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

Demonstration of Subscale Cermet Fuel Specimen Fabrication Approach Using Spark Plasma Sintering and Diffusion Bonding

NASA Technical Reports Server (NTRS)

Barnes, Marvin W.; Tucker, Dennis S.; Benensky, Kelsa M.

2018-01-01

Nuclear thermal propulsion (NTP) has the potential to expand the limits of human space exploration by enabling crewed missions to Mars and beyond. The viability of NTP hinges on the development of a robust nuclear fuel material that can perform in the harsh operating environment (> or = 2500K, reactive hydrogen) of a nuclear thermal rocket (NTR) engine. Efforts are ongoing to develop fuel material and to assemble fuel elements that will be stable during the service life of an NTR. Ceramic-metal (cermet) fuels are being actively pursued by NASA Marshall Space Flight Center (MSFC) due to their demonstrated high-temperature stability and hydrogen compatibility. Building on past cermet fuel development research, experiments were conducted to investigate a modern fabrication approach for cermet fuel elements. The experiments used consolidated tungsten (W)-60vol%zirconia (ZrO2) compacts that were formed via spark plasma sintering (SPS). The consolidated compacts were stacked and diffusion bonded to assess the integrity of the bond lines and internal cooling channel cladding. The assessment included hot hydrogen testing of the manufactured surrogate fuel and pure W for 45 minutes at 2500 K in the compact fuel element environmental test (CFEET) system. Performance of bonded W-ZrO2 rods was compared to bonded pure W rods to access bond line integrity and composite stability. Bonded surrogate fuels retained structural integrity throughout testing and incurred minimal mass loss.

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.

Research on hypersonic aircraft using pre-cooled turbojet engines

NASA Astrophysics Data System (ADS)

Taguchi, Hideyuki; Kobayashi, Hiroaki; Kojima, Takayuki; Ueno, Atsushi; Imamura, Shunsuke; Hongoh, Motoyuki; Harada, Kenya

2012-04-01

Systems analysis of a Mach 5 class hypersonic aircraft is performed. The aircraft can fly across the Pacific Ocean in 2 h. A multidisciplinary optimization program for aerodynamics, structure, propulsion, and trajectory is used in the analysis. The result of each element model is improved using higher accuracy analysis tools. The aerodynamic performance of the hypersonic aircraft is examined through hypersonic wind tunnel tests. A thermal management system based on the data of the wind tunnel tests is proposed. A pre-cooled turbojet engine is adopted as the propulsion system for the hypersonic aircraft. The engine can be operated continuously from take-off to Mach 5. This engine uses a pre-cooling cycle using cryogenic liquid hydrogen. The high temperature inlet air of hypersonic flight would be cooled by the same liquid hydrogen used as fuel. The engine is tested under sea level static conditions. The engine is installed on a flight test vehicle. Both liquid hydrogen fuel and gaseous hydrogen fuel are supplied to the engine from a tank and cylinders installed within the vehicle. The designed operation of major components of the engine is confirmed. A large amount of liquid hydrogen is supplied to the pre-cooler in order to make its performance sufficient for Mach 5 flight. Thus, fuel rich combustion is adopted at the afterburner. The experiments are carried out under the conditions that the engine is mounted upon an experimental airframe with both set up either horizontally or vertically. As a result, the operating procedure of the pre-cooled turbojet engine is demonstrated.

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

Mohanty, Subhasish; Majumdar, Saurindranath

Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

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

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

Buden, D.

1993-08-11

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

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

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

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

NASA Technical Reports Server (NTRS)

Clark, John S.; Walton, James T.; Mcguire, Melissa L.

1992-01-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

NASA Astrophysics Data System (ADS)

Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

2017-07-01

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

Fuel injection assembly for gas turbine engine combustor

NASA Technical Reports Server (NTRS)

Candy, Anthony J. (Inventor); Glynn, Christopher C. (Inventor); Barrett, John E. (Inventor)

2002-01-01

A fuel injection assembly for a gas turbine engine combustor, including at least one fuel stem, a plurality of concentrically disposed tubes positioned within each fuel stem, wherein a cooling supply flow passage, a cooling return flow passage, and a tip fuel flow passage are defined thereby, and at least one fuel tip assembly connected to each fuel stem so as to be in flow communication with the flow passages, wherein an active cooling circuit for each fuel stem and fuel tip assembly is maintained by providing all active fuel through the cooling supply flow passage and the cooling return flow passage during each stage of combustor operation. The fuel flowing through the active cooling circuit is then collected so that a predetermined portion thereof is provided to the tip fuel flow passage for injection by the fuel tip assembly.

Fuel supply device for supplying fuel to an engine combustor

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

Lindsay, M.H.; Kerr, W.B.

1990-05-29

This patent describes a variable flow rate fuel supply device for supplying fuel to an engine combustor. It comprises: fuel metering means having a fuel valve means for controlling the flow rate of fuel to the combustor; piston means for dividing a first cooling fluid chamber from a second cooling fluid chamber; coupling means for coupling the piston means to the fuel valve means; and cooling fluid supply means in communication with the first and second cooling fluid chamber for producing a first pressure differential across the piston means for actuating the fuel valve means in a first direction, andmore » for producing a second pressure differential across the piston means for actuating the valve means in a second direction opposite the first direction, to control the flow rate of the fuel through the fuel metering means and into the engine combustor; and means for positioning the fuel metering means within the second cooling air chamber enabling the cooling air supply means to both cool the fuel metering means and control the fuel supply rate of fuel supplied by the fuel metering means to the combustor.« less

High-density fuel combustion and cooling investigation. [kengine design

NASA Technical Reports Server (NTRS)

Labotz, R. J.; Rousar, D. C.; Valler, H. W.

1980-01-01

The analysis, design, fabrication and testing of several engine configurations are discussed with respect to the combustion and heat transfer characteristics of LOX/RP-1 at chamber pressures between 6895 and 13790 kPa (1000 and 2000 psia). The different engine configurations discussed include: 8274 kPa and 13790 kPa (1200 psia and 2000 psia) chamber pressure injectors with like doublet and preatomized triplet elements; cooled and uncooled acoustic resonators; and graphite, regeneratively cooled and calorimetric chambers ranging in length from 27.9 to 37.5 cm (11 to 15 in.). A high pressure LOX/RP-1 spark igniter is also evaluated.

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

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

Yang, Won Sik; Lin, C. S.; Hader, J. S.

2016-01-30

This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the dischargedmore » fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements« less

WATER BOILER REACTOR

DOEpatents

King, L.D.P.

1960-11-22

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

The Benefits of Nuclear Thermal Propulsion (NTP) in an Evolvable Mars Campaign

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Mccurdy, David R.

2014-01-01

NTR: High thrust high specific impulse (2 x LOXLH2chemical) engine uses high power density fission reactor with enriched uranium fuel as thermal power source. Reactor heat is removed using H2propellant which is then exhausted to produce thrust. Conventional chemical engine LH2tanks, turbopumps, regenerative nozzles and radiation-cooled shirt extensions used --NTR is next evolutionary step in high performance liquid rocket engines During the Rover program, a common fuel element tie tube design was developed and used in the design of the 50 klbf Kiwi-B4E (1964), 75 klbf Phoebus-1B (1967), 250 klbf Phoebus-2A (June 1968), then back down to the 25 klbf Pewee engine (Nov-Dec 1968) NASA and DOE are using this same approach: design, build, ground then flight test a small engine using a common fuel element that is scalable to a larger 25 klbf thrust engine needed for human missions

Increase in the efficiency of a high-speed ramjet on hydrocarbon fuel at the flying vehicle acceleration up to M = 6+

NASA Astrophysics Data System (ADS)

Abashev, V. M.; Korabelnikov, A. V.; Kuranov, A. L.; Tretyakov, P. K.

2017-10-01

At the analysis of the work process in a ramjet, a complex consideration of the ensemble of problems the solution of which determines the engine efficiency appears reasonable. The main problems are ensuring a high completeness of fuel combustion and minimal hydraulic losses, the reliability of cooling of high-heat areas with the use of the fuel cooling resource, and ensuring the strength of the engine duct elements under non-uniform heat loads due to fuel combustion in complex gas-dynamic flow structures. The fundamental techniques and approaches to the solution of above-noted problems are considered in the present report, their novelty and advantages in comparison with conventional techniques are substantiated. In particular, a technique of the arrangement of an intense (pre-detonation) combustion regime for ensuring a high completeness of fuel combustion and minimal hydraulic losses at a smooth deceleration of a supersonic flow down to the sound velocity using the pulsed-periodic gas-dynamic flow control has been proposed. A technique has been proposed for cooling the high-heat areas, which employs the cooling resource of the hydrocarbon fuel, including the process of the kerosene chemical transformation (conversion) using the nano-catalysts. An analysis has shown that the highly heated structure will operate in the elastic-plastic domain of the behavior of constructional materials, which is directly connected to the engine operation resource. There arise the problems of reducing the ramjet shells depending on deformations. The deformations also lead to a significant influence on the work process in the combustor and, naturally, on the heat transfer process and the performance of catalysts (the action of plastic and elastic deformations of restrained shells). The work presents some results illustrating the presence of identified problems. A conclusion is drawn about the necessity of formulating a complex investigation both with the realization in model experiments and execution of computational and theoretical investigations.

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

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

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

14 CFR Appendix A to Part 43 - Major Alterations, Major Repairs, and Preventive Maintenance

Code of Federal Regulations, 2013 CFR

2013-01-01

... hub design. (iii) Changes in the governor or control design. (iv) Installation of a propeller governor.... (iv) Engine mounts. (v) Control system. (vi) Landing gear. (vii) Hull or floats. (viii) Elements of an... of gravity limits of the aircraft. (xii) Changes to the basic design of the fuel, oil, cooling...

Application of numerical methods to heat transfer and thermal stress analysis of aerospace vehicles

NASA Technical Reports Server (NTRS)

Wieting, A. R.

1979-01-01

The paper describes a thermal-structural design analysis study of a fuel-injection strut for a hydrogen-cooled scramjet engine for a supersonic transport, utilizing finite-element methodology. Applications of finite-element and finite-difference codes to the thermal-structural design-analysis of space transports and structures are discussed. The interaction between the thermal and structural analyses has led to development of finite-element thermal methodology to improve the integration between these two disciplines. The integrated thermal-structural analysis capability developed within the framework of a computer code is outlined.

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

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

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

2013-07-01

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

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

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

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

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

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

Burners and combustion apparatus for carbon nanomaterial production

DOEpatents

Alford, J. Michael; Diener, Michael D; Nabity, James; Karpuk, Michael

2013-02-05

The invention provides improved burners, combustion apparatus, and methods for carbon nanomaterial production. The burners of the invention provide sooting flames of fuel and oxidizing gases. The condensable products of combustion produced by the burners of this invention produce carbon nanomaterials including without limitation, soot, fullerenic soot, and fullerenes. The burners of the invention do not require premixing of the fuel and oxidizing gases and are suitable for use with low vapor pressure fuels such as those containing substantial amounts of polyaromatic hydrocarbons. The burners of the invention can operate with a hot (e.g., uncooled) burner surface and require little, if any, cooling or other forms of heat sinking. The burners of the invention comprise one or more refractory elements forming the outlet of the burner at which a flame can be established. The burners of the invention provide for improved flame stability, can be employed with a wider range of fuel/oxidizer (e.g., air) ratios and a wider range of gas velocities, and are generally more efficient than burners using water-cooled metal burner plates. The burners of the invention can also be operated to reduce the formation of undesirable soot deposits on the burner and on surfaces downstream of the burner.

Burners and combustion apparatus for carbon nanomaterial production

DOEpatents

Alford, J. Michael; Diener, Michael D.; Nabity, James; Karpuk, Michael

2007-10-09

The invention provides improved burners, combustion apparatus, and methods for carbon nanomaterial production. The burners of the invention provide sooting flames of fuel and oxidizing gases. The condensable products of combustion produced by the burners of this invention produce carbon nanomaterials including without limitation, soot, fullerenic soot, and fullerenes. The burners of the invention do not require premixing of the fuel and oxidizing gases and are suitable for use with low vapor pressure fuels such as those containing substantial amounts of polyaromatic hydrocarbons. The burners of the invention can operate with a hot (e.g., uncooled) burner surface and require little, if any, cooling or other forms of heat sinking. The burners of the invention comprise one or more refractory elements forming the outlet of the burner at which a flame can be established. The burners of the invention provide for improved flame stability, can be employed with a wider range of fuel/oxidizer (e.g., air) ratios and a wider range of gas velocities, and are generally more efficient than burners using water-cooled metal burner plates. The burners of the invention can also be operated to reduce the formation of undesirable soot deposits on the burner and on surfaces downstream of the burner.

Regeneratively cooled rocket engine for space storable propellants

NASA Technical Reports Server (NTRS)

Wagner, W. R.; Waldman, B. J.

1973-01-01

Analyses and experimental studies were performed with the OF2 (F2/O2)/B2H6 propellant combination over a range in operating conditions to determine suitability for a space storable pressure fed engine configuration for an extended flight space vehicle configuration. The regenerative cooling mode selected for the thrust chamber was explored in detail with the use of both the fuel and oxidizer as coolants in an advanced milled channel construction thrust chamber design operating at 100 psia chamber pressure and a nominal mixture ratio of 3.0 with a 60:1 area ratio nozzle. Benefits of the simultaneous cooling as related to gaseous injection of both fuel and oxidizer propellants were defined. Heat transfer rates, performance and combustor stability were developed for impinging element triplet injectors in uncooled copper calorimeter hardware with flow, pressure and temperature instrumentation. Evaluation of the capabilities of the B2H6 and OF2 during analytical studies and numerous tests with flow through electrically heated blocks provided design criteria for subsequent regenerative chamber design and fabrication.

NEUTRONIC REACTORS

DOEpatents

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

1958-10-14

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

NASA Astrophysics Data System (ADS)

Dautray, Robert

2011-06-01

The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation, etc., radioprotection measures and treatment for the "transuranic" elements. For a long period of time, France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts and this could per se constitute one of the national priorities. Nous sommes naturellement bien conscients de l'énorme problème qui se pose au Japon actuellement comme suite au tremblement de terre et au tsunami de mars 2011 et leurs conséquences, notamment sur des installations électronucléaires. Le texte que nous présentons concerne des conditions totalement générales, indépendantes des problèmes spécifiques de sûreté qu'il faudra, de toute façon, traiter dans le cadre d'un développement éventuel de l'énergie nucléaire.We are aware, of course, of the huge problem that Japan has to deal with the aftermath of the quake and tsunami of March 2011 and their consequences on electronuclear power plants. The text that we present here concerns general physical topics independent of the specific safety problems, general physical topics which will have to be solved in the case of a contingent development of electronuclear power plants.

Initial verification and validation of RAZORBACK - A research reactor transient analysis code

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

Talley, Darren G.

2015-09-01

This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actualmore » ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.« less

Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

NASA Technical Reports Server (NTRS)

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

2015-01-01

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

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

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

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

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

Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

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

Gryzinski, M.A.; Wielgosz, M.

The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel ismore » additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with self-cooling system (enhanced comparing to the cooling properties inherent with regular rector pool water flows), -proper cooling conditions can be ensured by an appropriate water flow, so the resistance to flow has to be optimised, -the requirement of the minimum resistance to water flow leads to the openwork design of the fuel element separator, which, on the other hand, has to be strong enough to ensure the needed strength for mechanical load due to the fuel weight and forces associated with the water flow, -the possibility of changing beam and flux qualities by rotating the converter or repositioning the converter plates by moving or replacing with another materials. In order to minimize the neutron activation of the fuel in the converter, the possibility was predicted to remove the converter and to replace it with an aluminium dummy for the time when the beam at the channel H2 is not used. This means that both, the converter and the dummy, have to be easily removable from the converter socket. There has to be also the place in the water pool, near the research stand or in technological pool, where the converter can be safely stored (this place have to be proper for operation with plates i.e. changing amount of plates). Thermal and neutron load of the fuel plates in the converter will be inhomogeneous. In order to equalize these loads, the converter should be designed in such way that it would be possible to change the order of fuel plates. Moreover replacing the amount of the plates gives the opportunity to obtain different fluxes of neutrons (quantitatively and qualitatively i.e. energetically). The project of the converter is based on Monte Carlo calculation concerning neutron production and on Computational Fluid Dynamics (CFD) i.e. modelling of converter for thermodynamical aspects. (authors)« less

Development and Experimental Evaluation of Passive Fuel Cell Thermal Control

NASA Technical Reports Server (NTRS)

Colozza, Anthony J.; Jakupca, Ian J.; Castle, Charles H.; Burke, Kenneth A.

2014-01-01

To provide uniform cooling for a fuel cell stack, a cooling plate concept was evaluated. This concept utilized thin cooling plates to extract heat from the interior of a fuel cell stack and move this heat to a cooling manifold where it can be transferred to an external cooling fluid. The advantages of this cooling approach include a reduced number of ancillary components and the ability to directly utilize an external cooling fluid loop for cooling the fuel cell stack. A number of different types of cooling plates and manifolds were developed. The cooling plates consisted of two main types; a plate based on thermopyrolytic graphite (TPG) and a planar (or flat plate) heat pipe. The plates, along with solid metal control samples, were tested for both thermal and electrical conductivity. To transfer heat from the cooling plates to the cooling fluid, a number of manifold designs utilizing various materials were devised, constructed, and tested. A key aspect of the manifold was that it had to be electrically nonconductive so it would not short out the fuel cell stack during operation. Different manifold and cooling plate configurations were tested in a vacuum chamber to minimize convective heat losses. Cooling plates were placed in the grooves within the manifolds and heated with surface-mounted electric pad heaters. The plate temperature and its thermal distribution were recorded for all tested combinations of manifold cooling flow rates and heater power loads. This testing simulated the performance of the cooling plates and manifold within an operational fuel cell stack. Different types of control valves and control schemes were tested and evaluated based on their ability to maintain a constant temperature of the cooling plates. The control valves regulated the cooling fluid flow through the manifold, thereby controlling the heat flow to the cooling fluid. Through this work, a cooling plate and manifold system was developed that could maintain the cooling plates within a minimal temperature band with negligible thermal gradients over power profiles that would be experienced within an operating fuel cell stack.

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

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Air-Cooled Stack Freeze Tolerance Freeze Failure Modes and Freeze Tolerance Strategies for GenDriveTM Material Handling Application Systems and Stacks Final Scientific Report

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

Hancock, David, W.

2012-02-14

Air-cooled stack technology offers the potential for a simpler system architecture (versus liquid-cooled) for applications below 4 kilowatts. The combined cooling and cathode air allows for a reduction in part count and hence a lower cost solution. However, efficient heat rejection challenges escalate as power and ambient temperature increase. For applications in ambient temperatures below freezing, the air-cooled approach has additional challenges associated with not overcooling the fuel cell stack. The focus of this project was freeze tolerance while maintaining all other stack and system requirements. Through this project, Plug Power advanced the state of the art in technology formore » air-cooled PEM fuel cell stacks and related GenDrive material handling application fuel cell systems. This was accomplished through a collaborative work plan to improve freeze tolerance and mitigate freeze-thaw effect failure modes within innovative material handling equipment fuel cell systems designed for use in freezer forklift applications. Freeze tolerance remains an area where additional research and understanding can help fuel cells to become commercially viable. This project evaluated both stack level and system level solutions to improve fuel cell stack freeze tolerance. At this time, the most cost effective solutions are at the system level. The freeze mitigation strategies developed over the course of this project could be used to drive fuel cell commercialization. The fuel cell system studied in this project was Plug Power's commercially available GenDrive platform providing battery replacement for equipment in the material handling industry. The fuel cell stacks were Ballard's commercially available FCvelocity 9SSL (9SSL) liquid-cooled PEM fuel cell stack and FCvelocity 1020ACS (Mk1020) air-cooled PEM fuel cell stack.« less

Cooling System Design for PEM Fuel Cell Powered Air Vehicles

DTIC Science & Technology

2010-06-18

Research Laboratory (NRL) has developed a proton exchange membrane fuel cell ( PEMFC ) powered unmanned air vehicle (UAV) called the Ion Tiger. The Ion Tiger...to design a cooling system for the Ion Tiger and investigate cooling approaches that may be suitable for future PEMFC powered air vehicles. The...modifications) to other PEMFC systems utilizing a CHE for cooling. 18-06-2010 Memorandum Report Unmanned Air Vehicle UAV Fuel cell PEM Cooling Radiator January

Thermos reactors

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

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

Flame tolerant secondary fuel nozzle

DOEpatents

Khan, Abdul Rafey; Ziminsky, Willy Steve; Wu, Chunyang; Zuo, Baifang; Stevenson, Christian Xavier

2015-02-24

A combustor for a gas turbine engine includes a plurality of primary nozzles configured to diffuse or premix fuel into an air flow through the combustor; and a secondary nozzle configured to premix fuel with the air flow. Each premixing nozzle includes a center body, at least one vane, a burner tube provided around the center body, at least two cooling passages, a fuel cooling passage to cool surfaces of the center body and the at least one vane, and an air cooling passage to cool a wall of the burner tube. The cooling passages prevent the walls of the center body, the vane(s), and the burner tube from overheating during flame holding events.

Liquid metal cooled nuclear reactors with passive cooling system

DOEpatents

Hunsbedt, Anstein; Fanning, Alan W.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Variable area fuel cell cooling

DOEpatents

Kothmann, Richard E.

1982-01-01

A fuel cell arrangement having cooling fluid flow passages which vary in surface area from the inlet to the outlet of the passages. A smaller surface area is provided at the passage inlet, which increases toward the passage outlet, so as to provide more uniform cooling of the entire fuel cell. The cooling passages can also be spaced from one another in an uneven fashion.

A COOLED NEUTRONIC REACTOR

DOEpatents

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

1960-03-15

A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

Small hydrogen/oxygen rocket flowfield behavior from heat flux measurements

NASA Technical Reports Server (NTRS)

Reed, Brian D.

1993-01-01

The mixing and heat transfer phenomena in small rocket flow fields with fuel film cooling is not well understood. An instrumented, water-cooled chamber with a gaseous hydrogen/gaseous oxygen injector was used to gather steady-state inner and outer wall temperature profiles. The chamber was tested at 414 kPa (60 psia) chamber pressure, from mixture ratios of 3.41 to 8.36. Sixty percent of the fuel was used for film cooling. These temperature profiles were used as boundary conditions in a finite element analysis program, MSC/NASTRAN, to calculate the local radial and axial heat fluxes in the chamber wall. The normal heat fluxes were then calculated and used as a diagnostic of the rocket's flow field behavior. The normal heat fluxes determined were on the order of 1.0 to 3.0 MW/meters squared (0.6 to 1.8 Btu/sec-inches squared). In the cases where mixture ratio was 5 or above, there was a sharp local heat flux maximum in the barrel section of the chamber. This local maximum seems to indicate a reduction or breakdown of the fuel film cooling layer, possibly due to increased mixing in the shear layer between the film and core flows. However, the flow was thought to be completely laminar, as the throat Reynolds numbers were below 50,000 for all the cases. The increased mixing in the shear layer in the higher mixture ratio cases appeared not to be due to the transition of the flow from laminar to turbulent, but rather due to increased reactions between the hydrogen film and oxidizer-rich core flows.

Catalytic combustor for integrated gasification combined cycle power plant

DOEpatents

Bachovchin, Dennis M [Mauldin, SC; Lippert, Thomas E [Murrysville, PA

2008-12-16

A gasification power plant 10 includes a compressor 32 producing a compressed air flow 36, an air separation unit 22 producing a nitrogen flow 44, a gasifier 14 producing a primary fuel flow 28 and a secondary fuel source 60 providing a secondary fuel flow 62 The plant also includes a catalytic combustor 12 combining the nitrogen flow and a combustor portion 38 of the compressed air flow to form a diluted air flow 39 and combining at least one of the primary fuel flow and secondary fuel flow and a mixer portion 78 of the diluted air flow to produce a combustible mixture 80. A catalytic element 64 of the combustor 12 separately receives the combustible mixture and a backside cooling portion 84 of the diluted air flow and allows the mixture and the heated flow to produce a hot combustion gas 46 provided to a turbine 48. When fueled with the secondary fuel flow, nitrogen is not combined with the combustor portion.

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

NASA Astrophysics Data System (ADS)

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

2009-03-01

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

Heat pipe nuclear reactor for space power

NASA Technical Reports Server (NTRS)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

Syngas Production By Thermochemical Conversion Of H2o And Co2 Mixtures Using A Novel Reactor Design

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

Pearlman, Howard; Chen, Chien-Hua

The Department of Energy awarded Advanced Cooling Technologies, Inc. (ACT) an SBIR Phase II contract (#DE-SC0004729) to develop a high-temperature solar thermochemical reactor for syngas production using water and/or carbon dioxide as feedstocks. The technology aims to provide a renewable and sustainable alternative to fossil fuels, promote energy independence and mitigate adverse issues associated with climate change by essentially recycling carbon from carbon dioxide emitted by the combustion of hydrocarbon fuels. To commercialize the technology and drive down the cost of solar fuels, new advances are needed in materials development and reactor design, both of which are integral elements inmore » this program.« less

Sodium aluminum-iron phosphate glass-ceramics for immobilization of lanthanide oxide wastes from pyrochemical reprocessing of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Stefanovsky, S. V.; Stefanovsky, O. I.; Kadyko, M. I.; Nikonov, B. S.

2018-03-01

Sodium aluminum (iron) phosphate glass ceramics containing of up to 20 wt.% rare earth (RE) oxides simulating pyroprocessing waste were produced by melting at 1250 °C followed by either quenching or slow cooling to room temperature. The iron-free glass-ceramics were composed of major glass and minor phosphotridymite and monazite. The iron-bearing glass-ceramics were composed of major glass and minor monazite and Na-Al-Fe orthophosphate at low waste loadings (5-10 wt.%) and major orthophosphate and minor monazite as well as interstitial glass at high waste loadings (15-20 wt.%). Slowly cooled samples contained higher amount of crystalline phases than quenched ones. Monazite is major phase for REs. Leach rates from the materials of major elements (Na, Al, Fe, P) are 10-5-10-7 g cm-2 d-1, RE elements - lower than 10-5 g cm-2 d-1.

TERNARY ALLOY-CONTAINING PLUTONIUM

DOEpatents

Waber, J.T.

1960-02-23

Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

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

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

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

2016-05-01

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

Flame holding tolerant fuel and air premixer for a gas turbine combustor

DOEpatents

York, William David; Johnson, Thomas Edward; Ziminsky, Willy Steve

2012-11-20

A fuel nozzle with active cooling is provided. It includes an outer peripheral wall, a nozzle center body concentrically disposed within the outer wall in a fuel and air pre-mixture. The fuel and air pre-mixture includes an air inlet, a fuel inlet and a premixing passage defined between the outer wall in the center body. A gas fuel flow passage is provided. A first cooling passage is included within the center body in a second cooling passage is defined between the center body and the outer wall.

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.

Thermal protection performance of opposing jet generating with solid fuel

NASA Astrophysics Data System (ADS)

Shen, Binxian; Liu, Weiqiang

2018-03-01

A light and small gas supply device, which uses fuel gas generating with solid fuel as coolant gas, is introduced for opposing jet thermal protection in hypersonic vehicles. A numerical study on heat flux reduction in hypersonic flow with opposing jet is conducted to investigate the cooling efficiency of fuel gas. Flow field and cooling efficiency at different jet temperatures, as well as the effect of fuel gas, are determined. Detailed results show that shock stand-off distance changes with an increase in jet pressure ratio and remains constant with an increase in jet temperature. Cooling efficiency weakens with an increase in jet temperature and can be strengthened by enhancing jet pressure. Lastly, a remarkable heat flux reduction is observed with fuel gas injection with respect to no fuel gas injection when jet temperature reaches 900 K, thereby proving the positive cooling efficiency of fuel gas.

Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

NASA Technical Reports Server (NTRS)

Klann, P. G.; Lantz, E.

1973-01-01

A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

Demonstration of Passive Fuel Cell Thermal Management Technology

NASA Technical Reports Server (NTRS)

Burke, Kenneth A.; Jakupca, Ian; Colozza, Anthony; Wynne, Robert; Miller, Michael; Meyer, Al; Smith, William

2012-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates and integrated heat exchanger technology to collect the heat from the cooling plates (Ref. 1). The next step in the development of this passive thermal approach was the demonstration of the control of the heat removal process and the demonstration of the passive thermal control technology in actual fuel cell stacks. Tests were run with a simulated fuel cell stack passive thermal management system outfitted with passive cooling plates, an integrated heat exchanger and two types of cooling flow control valves. The tests were run to demonstrate the controllability of the passive thermal control approach. Finally, successful demonstrations of passive thermal control technology were conducted with fuel cell stacks from two fuel cell stack vendors.

ETR, TRA642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF ...

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

ETR, TRA-642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF FLOOR AND FACES NORTH. OUTER WALL OF STORAGE CANAL IS AT RIGHT. SHIELDING IS THICKER AT LOWER LEVEL, WHERE SPENT FUEL ELEMENTS WILL COOL AFTER REMOVAL FROM REACTOR. INL NEGATIVE NO. 56-1401. Jack L. Anderson, Photographer, 5/1/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Nonlinear heat transfer and structural analyses of SSME turbine blades

NASA Technical Reports Server (NTRS)

Abdul-Aziz, A.; Kaufman, A.

1987-01-01

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

Fuel processor and method for generating hydrogen for fuel cells

DOEpatents

Ahmed, Shabbir [Naperville, IL; Lee, Sheldon H. D. [Willowbrook, IL; Carter, John David [Bolingbrook, IL; Krumpelt, Michael [Naperville, IL; Myers, Deborah J [Lisle, IL

2009-07-21

A method of producing a H.sub.2 rich gas stream includes supplying an O.sub.2 rich gas, steam, and fuel to an inner reforming zone of a fuel processor that includes a partial oxidation catalyst and a steam reforming catalyst or a combined partial oxidation and stream reforming catalyst. The method also includes contacting the O.sub.2 rich gas, steam, and fuel with the partial oxidation catalyst and the steam reforming catalyst or the combined partial oxidation and stream reforming catalyst in the inner reforming zone to generate a hot reformate stream. The method still further includes cooling the hot reformate stream in a cooling zone to produce a cooled reformate stream. Additionally, the method includes removing sulfur-containing compounds from the cooled reformate stream by contacting the cooled reformate stream with a sulfur removal agent. The method still further includes contacting the cooled reformate stream with a catalyst that converts water and carbon monoxide to carbon dioxide and H.sub.2 in a water-gas-shift zone to produce a final reformate stream in the fuel processor.

Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

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

Durkee, Jr., Joe W.

A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20,more » 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/ 137Cs 134Cs/ 154Eu, and 154Eu/ 137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the FA models serve as source materials for the pre- and postelectrorefining models to be reported in Parts 2 and 3.« less

METHOD OF SUSTAINING A NEUTRONIC CHAIN REACTING SYSTEM

DOEpatents

Fermi, E.; Leverett, M.C.

1957-11-12

This patent relates to neutronic reactors and a method of sustainlng a chain reaction. The reactor shown in the patent for carrying out the method is the gas-cooled type comprised of a solid moderator having a plurality of passages therethrough for receiving bodies of fissionable material. In carrying out the method, the reactor is loaded by inserting in the passages fuel elements and moderator material in a proportion to sustain a chain reaction As the reproduction ratio decreases below the desired fiiaire due to impurities formed during operation of the reactor, the moderator material is gradually replaced with additional fuel material to maintain the reproduction ratio above unity.

Experimental investigations on active cooling thermal protection structure of hydrocarbon-fueled scramjet combustor in arc heated facility

NASA Astrophysics Data System (ADS)

Jianqiang, Tu; Jinlong, Peng; Xianning, Yang; Lianzhong, Chen

2016-10-01

The active cooling thermal protection technology is the efficient method to resolve the long-duration work and reusable problems of hydrocarbon-fueled scramjet combustor, where worst thermo-mechanical loads occur. The fuel is passed through coolant channels adjacent to the heated surfaces to absorb heat from the heating exchanger panels, prior to injection into the combustor. The heating exchanger both cooled down the wall temperature of the combustor wall and heats and cracks the hydrocarbon fuel inside the panel to permit an easier combustion and satisfying combustion efficiency. The subscale active cooling metallic panels, with dimensions of 100×100 mm and different coolant channel sizes, have been tested under typical combustion thermal environment produced by arc heated Turbulent Flow Duct (TFD). The heat exchange ability of different coolant channel sizes has been obtained. The big-scale active cooling metallic panel, with dimensions of 100 × 750 mm and the coolant channel sizes of better heating exchange performance, has been made and tested in the big-scale arc heated TFD facility. The test results show that the local superheated ablation is easy to happen for the cooling fuel assigned asymmetrically in the bigscale active cooling metallic panel, and the cooling fuel rate can reduce 8%˜10% after spraying the Thermal Barrier Coating (TBC) in the heating surface.

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

PubMed

Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

2018-05-01

Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

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.

Preliminary neutronic analysis of a cavity test reactor

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

A reference configuration was calculated for a cavity test reactor to be used for testing the gascore nuclear rocket concept. A thermal flux of 4.1 x 10 to the 14th power neutrons per square centimeter per second in the cavity was provided by a driver fuel loading of 6.4 kg of enriched uranium in MTR fuel elements. The reactor was moderated and cooled by heavy water and reflected with 25.4 cm of beryllium. Power generation of 41.3 MW in the driver fuel is rejected to a heat sink. Design effort was directed toward minimization of driver power while maintaining 2.7 MW in the cavity during a test run. Ancillary data on material reactivity worths, reactivity coefficients, flux spectra, and power distributions are reported.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

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

Oil cooling system for a gas turbine engine

NASA Technical Reports Server (NTRS)

Coffinberry, G. A.; Kast, H. B. (Inventor)

1977-01-01

A gas turbine engine fuel delivery and control system is provided with means to recirculate all fuel in excess of fuel control requirements back to aircraft fuel tank, thereby increasing the fuel pump heat sink and decreasing the pump temperature rise without the addition of valving other than that normally employed. A fuel/oil heat exchanger and associated circuitry is provided to maintain the hot engine oil in heat exchange relationship with the cool engine fuel. Where anti-icing of the fuel filter is required, means are provided to maintain the fuel temperature entering the filter at or above a minimum level to prevent freezing thereof. Fluid circuitry is provided to route hot engine oil through a plurality of heat exchangers disposed within the system to provide for selective cooling of the oil.

Modeling of Fuel Film Cooling Using Steady State RANS and Unsteady DES Approaches

DTIC Science & Technology

2016-07-27

Briefing Charts 3. DATES COVERED (From - To) 21 July 2016 – 31 August 2016 4. TITLE AND SUBTITLE Modeling of Fuel Film Cooling Using Steady State RANS...Prescribed by ANSI Std. 239.18 1 Distribution A: Approved for Public Release; Distribution Unlimited. PA# 16391. Modeling of Fuel  Film  Cooling Using...Distribution Unlimited. PA# 16391. 3 Introduction • Fuel  film  cooling is critical for high performing boost engines  using the Oxygen Rich Staged

Development of Passive Fuel Cell Thermal Management Heat Exchanger

NASA Technical Reports Server (NTRS)

Burke, Kenneth A.; Jakupca, Ian J.; Colozza, Anthony J.

2010-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates that could conduct the heat, provide a sufficiently uniform temperature heat sink for each cell of the fuel cell stack, and be substantially lighter than the conventional thermal management approach. Tests were run with different materials to evaluate the design approach to a heat exchanger that could interface with the edges of the passive cooling plates. Measurements were made during fuel cell operation to determine the temperature of individual cooling plates and also to determine the temperature uniformity from one cooling plate to another.

Advanced Fuels for Chemical Lasers

DTIC Science & Technology

1975-05-01

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

Propulsion system needs

NASA Technical Reports Server (NTRS)

Gunn, Stanley

1991-01-01

The needs of the designer of a solid core nuclear rocket engine are discussed. Some of the topics covered include: (1) a flight thrust module/feed system module assembly; (2) a nuclear thermal rocket (NTR), expander cycle, dual T/P; (3) turbopump operating conditions; (4) typical system parameters; (5) growth capability composite fuel elements; (6) a NTR radiation cooled nozzle extension; (7) a NFS-3B Feed System; and (8) a NTR Integrated Pneumatic-Fluidics Control System.

Air-cooled, hydrogen-air fuel cell

NASA Technical Reports Server (NTRS)

Shelekhin, Alexander B. (Inventor); Bushnell, Calvin L. (Inventor); Pien, Michael S. (Inventor)

1999-01-01

An air-cooled, hydrogen-air solid polymer electrolyte (SPE) fuel cell with a membrane electrode assembly operatively associated with a fluid flow plate having at least one plate cooling channel extending through the plate and at least one air distribution hole extending from a surface of the cathode flow field into the plate cooling channel.

Convective Heat Transfer with and without Film Cooling in High Temperature, Fuel Rich and Lean Environments

NASA Astrophysics Data System (ADS)

Greiner, Nathan J.

Modern turbine engines require high turbine inlet temperatures and pressures to maximize thermal efficiency. Increasing the turbine inlet temperature drives higher heat loads on the turbine surfaces. In addition, increasing pressure ratio increases the turbine coolant temperature such that the ability to remove heat decreases. As a result, highly effective external film cooling is required to reduce the heat transfer to turbine surfaces. Testing of film cooling on engine hardware at engine temperatures and pressures can be exceedingly difficult and expensive. Thus, modern studies of film cooling are often performed at near ambient conditions. However, these studies are missing an important aspect in their characterization of film cooling effectiveness. Namely, they do not model effect of thermal property variations that occur within the boundary and film cooling layers at engine conditions. Also, turbine surfaces can experience significant radiative heat transfer that is not trivial to estimate analytically. The present research first computationally examines the effect of large temperature variations on a turbulent boundary layer. Subsequently, a method to model the effect of large temperature variations within a turbulent boundary layer in an environment coupled with significant radiative heat transfer is proposed and experimentally validated. Next, a method to scale turbine cooling from ambient to engine conditions via non-dimensional matching is developed computationally and the experimentally validated at combustion temperatures. Increasing engine efficiency and thrust to weight ratio demands have driven increased combustor fuel-air ratios. Increased fuel-air ratios increase the possibility of unburned fuel species entering the turbine. Alternatively, advanced ultra-compact combustor designs have been proposed to decrease combustor length, increase thrust, or generate power for directed energy weapons. However, the ultra-compact combustor design requires a film cooled vane within the combustor. In both these environments, the unburned fuel in the core flow encounters the oxidizer rich film cooling stream, combusts, and can locally heat the turbine surface rather than the intended cooling of the surface. Accordingly, a method to quantify film cooling performance in a fuel rich environment is prescribed. Finally, a method to film cool in a fuel rich environment is experimentally demonstrated.

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

NASA Technical Reports Server (NTRS)

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

2002-01-01

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

Method for generating hydrogen for fuel cells

DOEpatents

Ahmed, Shabbir; Lee, Sheldon H. D.; Carter, John David; Krumpelt, Michael

2004-03-30

A method of producing a H.sub.2 rich gas stream includes supplying an O.sub.2 rich gas, steam, and fuel to an inner reforming zone of a fuel processor that includes a partial oxidation catalyst and a steam reforming catalyst or a combined partial oxidation and stream reforming catalyst. The method also includes contacting the O.sub.2 rich gas, steam, and fuel with the partial oxidation catalyst and the steam reforming catalyst or the combined partial oxidation and stream reforming catalyst in the inner reforming zone to generate a hot reformate stream. The method still further includes cooling the hot reformate stream in a cooling zone to produce a cooled reformate stream. Additionally, the method includes removing sulfur-containing compounds from the cooled reformate stream by contacting the cooled reformate stream with a sulfur removal agent. The method still further includes contacting the cooled reformate stream with a catalyst that converts water and carbon monoxide to carbon dioxide and H.sub.2 in a water-gas-shift zone to produce a final reformate stream in the fuel processor.

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.

Thermal Propulsion Capture System Heat Exchanger Design

NASA Technical Reports Server (NTRS)

Richard, Evan M.

2016-01-01

One of the biggest challenges of manned spaceflight beyond low earth orbit and the moon is harmful radiation that astronauts would be exposed to on their long journey to Mars and further destinations. Using nuclear energy has the potential to be a more effective means of propulsion compared to traditional chemical engines (higher specific impulse). An upper stage nuclear engine would allow astronauts to reach their destination faster and more fuel efficiently. Testing these engines poses engineering challenges due to the need to totally capture the engine exhaust. The Thermal Propulsion Capture System is a concept for cost effectively and safely testing Nuclear Thermal Engines. Nominally, hydrogen exhausted from the engine is not radioactive, but is treated as such in case of fuel element failure. The Thermal Propulsion Capture System involves injecting liquid oxygen to convert the hydrogen exhaust into steam. The steam is then cooled and condensed into liquid water to allow for storage. The Thermal Propulsion Capture System concept for ground testing of a nuclear powered engine involves capturing the engine exhaust to be cooled and condensed before being stored. The hydrogen exhaust is injected with liquid oxygen and burned to form steam. That steam must be cooled to saturation temperatures before being condensed into liquid water. A crossflow heat exchanger using water as a working fluid will be designed to accomplish this goal. Design a cross flow heat exchanger for the Thermal Propulsion Capture System testing which: Eliminates the need for water injection cooling, Cools steam from 5800 F to saturation temperature, and Is efficient and minimizes water requirement.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Evaluation of a Passive Heat Exchanger Based Cooling System for Fuel Cell Applications

NASA Technical Reports Server (NTRS)

Colozza, Anthony J.; Burke, Kenneth A.

2011-01-01

Fuel cell cooling is conventionally performed with an actively controlled, dedicated coolant loop that exchanges heat with a separate external cooling loop. To simplify this system the concept of directly cooling a fuel cell utilizing a coolant loop with a regenerative heat exchanger to preheat the coolant entering the fuel cell with the coolant exiting the fuel cell was analyzed. The preheating is necessary to minimize the temperature difference across the fuel cell stack. This type of coolant system would minimize the controls needed on the coolant loop and provide a mostly passive means of cooling the fuel cell. The results indicate that an operating temperature of near or greater than 70 C is achievable with a heat exchanger effectiveness of around 90 percent. Of the heat exchanger types evaluated with the same type of fluid on the hot and cold side, a counter flow type heat exchanger would be required which has the possibility of achieving the required effectiveness. The number of heat transfer units required by the heat exchanger would be around 9 or greater. Although the analysis indicates the concept is feasible, the heat exchanger design would need to be developed and optimized for a specific fuel cell operation in order to achieve the high effectiveness value required.

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

NASA Technical Reports Server (NTRS)

Abdul-Aziz, Ali; Thompson, Robert L.

1988-01-01

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

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

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

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Cell module and fuel conditioner development

NASA Technical Reports Server (NTRS)

Hoover, D. Q., Jr.

1982-01-01

The phosphoric acid fuel cell module (stack) development which culminated in an 80 cell air-cooled stack with separated gas cooling and treed cooling plates is described. The performance of the 80 cell stack was approx. 100 mV per cell higher than that attained during phase 1. The components and materials performed stably for over 8000 hours in a 5 cell stack. The conceptual design of a fuel conditioning system is described.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Some advantages of methane in an aircraft gas turbine

NASA Technical Reports Server (NTRS)

Graham, R. W.; Glassman, A. J.

1980-01-01

Liquid methane, which can be manufactured from any of the hydrocarbon sources such as coal, shale biomass, and organic waste considered as a petroleum replacement for aircraft fuels. A simple cycle analysis is carried out for a turboprop engine flying a Mach 0.8 and 10, 688 meters (35,000 ft.) altitude. Cycle performance comparisions are rendered for four cases in which the turbine cooling air is cooled or not cooled by the methane fuel. The advantages and disadvantages of involving the fuel in the turbine cooling system are discussed. Methane combustion characteristics are appreciably different from Jet A and will require different combustor designs. Although a number of similar difficult technical problems exist, a highly fuel efficient turboprop engine burning methane appear to be feasible.

The Effect of Valve Cooling upon Maximum Permissible Engine Output as Limited by Knock

NASA Technical Reports Server (NTRS)

Munger, Maurice; Wilsted, H D; Mulcahy, B A

1942-01-01

A Wright GR-1820-G200 cylinder was tested over a wide range of fuel-air ratios at maximum permissible power output as limited by knock with three different degrees of valve cooling. The valves used were stock valves (solid inlet valve and hollow sodium-cooled exhaust valve), hollow valves with no coolant, and hollow valves with flowing water as a coolant. Curves showing the variation in maximum permissible values of inlet-air pressure, indicated mean effective pressure, cylinder charge, and indicated specific fuel consumption with change in fuel-air ratio and valve cooling are shown. The use of valves cooled by a stream of water passing through their hollow interiors permitted indicated mean effective pressures 10 percent higher than the mean effective pressures permissible with stock valves when the engine was operated with fuel-air ratios from 0.055 to 0.065. Operation of the engine with lean mixtures with uncooled hollow valves resulted in power output below the output obtained with the stock valves. The data show an increase in maximum permissible indicated mean effective pressure due to cooling the valves, which averages only 2.1 percent with fuel-air ratios from 0.075 to 0.105.

Fuel injector for use in a gas turbine engine

DOEpatents

Wiebe, David J.

2012-10-09

A fuel injector in a combustor apparatus of a gas turbine engine. An outer wall of the injector defines an interior volume in which an intermediate wall is disposed. A first gap is formed between the outer wall and the intermediate wall. The intermediate wall defines an internal volume in which an inner wall is disposed. A second gap is formed between the intermediate wall and the inner wall. The second gap receives cooling fluid that cools the injector. The cooling fluid provides convective cooling to the intermediate wall as it flows within the second gap. The cooling fluid also flows through apertures in the intermediate wall into the first gap where it provides impingement cooling to the outer wall and provides convective cooling to the outer wall. The inner wall defines a passageway that delivers fuel into a liner downstream from a main combustion zone.

Development of Passive Fuel Cell Thermal Management Technology

NASA Technical Reports Server (NTRS)

Burke, Kenneth A.; Jakupca, Ian; Colozza, Anthony

2011-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. The passive thermal management system relies on heat conduction within the cooling plate to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack rather than using a pumped loop cooling system to convectively remove the heat. Using the passive approach eliminates the need for a coolant pump and other cooling loop components which reduces fuel cell system mass and improves overall system reliability. Previous analysis had identified that low density, ultra-high thermal conductivity materials would be needed for the cooling plates in order to achieve the desired reductions in mass and the highly uniform thermal heat sink for each cell within a fuel cell stack. A pyrolytic graphite material was identified and fabricated into a thin plate using different methods. Also a development project with Thermacore, Inc. resulted in a planar heat pipe. Thermal conductivity tests were done using these materials. The results indicated that lightweight passive fuel cell cooling is feasible.

Design, Fabrication, and Testing of an Auxiliary Cooling System for Jet Engines

NASA Technical Reports Server (NTRS)

Leamy, Kevin; Griffiths, Jim; Andersen, Paul; Joco, Fidel; Laski, Mark; Balser, Jeffrey (Technical Monitor)

2001-01-01

This report summarizes the technical effort of the Active Cooling for Enhanced Performance (ACEP) program sponsored by NASA. It covers the design, fabrication, and integrated systems testing of a jet engine auxiliary cooling system, or turbocooler, that significantly extends the use of conventional jet fuel as a heat sink. The turbocooler is designed to provide subcooled cooling air to the engine exhaust nozzle system or engine hot section. The turbocooler consists of three primary components: (1) a high-temperature air cycle machine driven by engine compressor discharge air, (2) a fuel/ air heat exchanger that transfers energy from the hot air to the fuel and uses a coating to mitigate fuel deposits, and (3) a high-temperature fuel injection system. The details of the turbocooler component designs and results of the integrated systems testing are documented. Industry Version-Data and information deemed subject to Limited Rights restrictions are omitted from this document.

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

NASA Technical Reports Server (NTRS)

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

2016-01-01

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

14 CFR 27.1305 - Powerplant instruments.

Code of Federal Regulations, 2013 CFR

2013-01-01

... with § 27.955. (n) A gas temperature indicator for each turbine engine. (o) Means to enable the pilot... temperature indicator, for each— (1) Air cooled engine; (2) Rotorcraft with cooling shutters; and (3... condition with respect to cooling. (c) A fuel pressure indicator, for each pump-fed engine. (d) A fuel...

14 CFR 27.1305 - Powerplant instruments.

Code of Federal Regulations, 2014 CFR

2014-01-01

... with § 27.955. (n) A gas temperature indicator for each turbine engine. (o) Means to enable the pilot... temperature indicator, for each— (1) Air cooled engine; (2) Rotorcraft with cooling shutters; and (3... condition with respect to cooling. (c) A fuel pressure indicator, for each pump-fed engine. (d) A fuel...

14 CFR 27.1305 - Powerplant instruments.

Code of Federal Regulations, 2011 CFR

2011-01-01

... with § 27.955. (n) A gas temperature indicator for each turbine engine. (o) Means to enable the pilot... temperature indicator, for each— (1) Air cooled engine; (2) Rotorcraft with cooling shutters; and (3... condition with respect to cooling. (c) A fuel pressure indicator, for each pump-fed engine. (d) A fuel...

14 CFR 27.1305 - Powerplant instruments.

Code of Federal Regulations, 2012 CFR

2012-01-01

... with § 27.955. (n) A gas temperature indicator for each turbine engine. (o) Means to enable the pilot... temperature indicator, for each— (1) Air cooled engine; (2) Rotorcraft with cooling shutters; and (3... condition with respect to cooling. (c) A fuel pressure indicator, for each pump-fed engine. (d) A fuel...

Relation of Hydrogen and Methane to Carbon Monoxide in Exhaust Gases from Internal-Combustion Engines

NASA Technical Reports Server (NTRS)

Gerrish, Harold C; Tessmann, Arthur M

1935-01-01

The relation of hydrogen and methane to carbon monoxide in the exhaust gases from internal-combustion engines operating on standard-grade aviation gasoline, fighting-grade aviation gasoline, hydrogenated safety fuel, laboratory diesel fuel, and auto diesel fuel was determined by analysis of the exhaust gases. Two liquid-cooled single-cylinder spark-ignition, one 9-cylinder radial air-cooled spark-ignition, and two liquid-cooled single-cylinder compression-ignition engines were used.

Nuclear light bulb

NASA Technical Reports Server (NTRS)

Latham, Tom

1991-01-01

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

Processing and microstructural characterisation of a UO2-based ceramic for disposal studies on spent AGR fuel

NASA Astrophysics Data System (ADS)

Hiezl, Z.; Hambley, D. I.; Padovani, C.; Lee, W. E.

2015-01-01

Preparation and characterisation of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Given the relatively small differences in radionuclide inventory expected over longer time periods, the SIMFuel studied in this work is expected to be also representative of spent fuel after significantly longer periods (e.g. 1000 years). Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. SIMFuel pellets were up to 92% dense and during the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE) O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF. Metallic precipitates are generally spherical and have submicron particle size (0.8 ± 0.7 μm). Spherical oxide precipitates in SIMFuel measured up to 30 μm in diameter, but no data were available in the public domain to compare this to AGR SNF. The grain size of actual AGR SNF (∼ 3-30 μm) is larger than that measured in AGR SIMFuel (∼ 2-5 μm).

Thermally efficient melting and fuel reforming for glass making

DOEpatents

Chen, Michael S.; Painter, Corning F.; Pastore, Steven P.; Roth, Gary S.; Winchester, David C.

1991-01-01

An integrated process for utilizing waste heat from a glass making furnace. The hot off-gas from the furnace is initially partially cooled, then fed to a reformer. In the reformer, the partially cooled off-gas is further cooled against a hydrocarbon which is thus reformed into a synthesis gas, which is then fed into the glass making furnace as a fuel. The further cooled off-gas is then recycled back to absorb the heat from the hot off-gas to perform the initial cooling.

Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant Conceptual Design Engineering Report (CDER)

NASA Technical Reports Server (NTRS)

1981-01-01

The reference conceptual design of the magnetohydrodynamic (MHD) Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD, is summarized. Main elements of the design, systems, and plant facilities are illustrated. System design descriptions are included for closed cycle cooling water, industrial gas systems, fuel oil, boiler flue gas, coal management, seed management, slag management, plant industrial waste, fire service water, oxidant supply, MHD power ventilating

Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant Conceptual Design Engineering Report (CDER)

NASA Astrophysics Data System (ADS)

1981-09-01

The reference conceptual design of the magnetohydrodynamic (MHD) Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD, is summarized. Main elements of the design, systems, and plant facilities are illustrated. System design descriptions are included for closed cycle cooling water, industrial gas systems, fuel oil, boiler flue gas, coal management, seed management, slag management, plant industrial waste, fire service water, oxidant supply, MHD power ventilating

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT ...

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

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT BREAKS IN CLADDING OF FUEL ELEMENTS. COUNT-RATE METER IN TOP PANEL INDICATES AMOUNT OF RADIOACTIVITY. LOWER PANELS SUPPLY POWER AND AMPLIFICATION OF SIGNALS GENERATED BY SCINTILLATION COUNTER/PHOTOMULTIPLIER TUBE COMBINATION IN RESPONSE TO RADIOACTIVITY IN A SAMPLE OF THE COOLING WATER. INL NEGATIVE NO. 56-771. Jack L. Anderson, Photographer, 3/15/1956. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PRODUCTION OF PURIFIED URANIUM

DOEpatents

Burris, L. Jr.; Knighton, J.B.; Feder, H.M.

1960-01-26

A pyrometallurgical method for processing nuclear reactor fuel elements containing uranium and fission products and for reducing uranium compound; to metallic uranium is reported. If the material proccssed is essentially metallic uranium, it is dissolved in zinc, the sulution is cooled to crystallize UZn/sub 9/ , and the UZn/sub 9/ is distilled to obtain uranium free of fission products. If the material processed is a uranium compound, the sollvent is an alloy of zinc and magnesium and the remaining steps are the same.

REACTOR CONTROL

DOEpatents

Ruano, W.J.

1957-12-10

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Design and Economic Analysis of a Heating/Absorption Cooling System Operating with Municipal Solid Waste Digester: A Case Study of Gazi University

NASA Astrophysics Data System (ADS)

Coşar, Gökhan; Pooyanfar, Mirparham; Amirabedin, Ehsan; Topal, Hüseyin

2013-12-01

Recovering energy from municipal solid waste (MSW) is one of the most important issues of energy management in developed countries. This raises even more interest as world fossil fuel reserves diminish and fuel prices rise. Being one of main processes of waste disposal, anaerobic digestion can be used as a means to reduce fossil fuel and electricity consumption as well as reducing emissions. With growing demand for cooling in Turkey, especially during warm seasons and considering the energy costs, utilizing heat-driven absorption cooling systems coupled with an anaerobic digester for local cooling purposes is a potentially interesting alternative for electricity driven compression cooling. The aim of this article is to study the viability of utilizing biogas obtained from MSW anaerobic digestion as the main fuel for heating facilities of Gazi University, Turkey and also the energy source for an absorption cooling system designed for the central library of the aforementioned campus. The results prove that the suggested system is sustainably and financially appealing and has the potential to replace the conventional electricity driven cooling systems with a reasonable net present worth; moreover, it can notably reduce carbon dioxide emissions.

Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

NASA Astrophysics Data System (ADS)

Kaplan, Alexis C.; Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P.; Flaska, Marek; Pozzi, Sara A.

2014-11-01

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

Dual-fuel, dual-mode rocket engine

NASA Technical Reports Server (NTRS)

Martin, James A. (Inventor)

1989-01-01

The invention relates to a dual fuel, dual mode rocket engine designed to improve the performance of earth-to-orbit vehicles. For any vehicle that operates from the earth's surface to earth orbit, it is advantageous to use two different fuels during its ascent. A high density impulse fuel, such as kerosene, is most efficient during the first half of the trajectory. A high specific impulse fuel, such as hydrogen, is most efficient during the second half of the trajectory. The invention allows both fuels to be used with a single rocket engine. It does so by adding a minimum number of state-of-the-art components to baseline single made rocket engines, and is therefore relatively easy to develop for near term applications. The novelty of this invention resides in the mixing of fuels before exhaust nozzle cooling. This allows all of the engine fuel to cool the exhaust nozzle, and allows the ratio of fuels used throughout the flight depend solely on performance requirements, not cooling requirements.

Modeling and Comparison of Options for the Disposal of Excess Weapons Plutonium in Russia

DTIC Science & Technology

2002-04-01

fuel LWR cooling time LWR Pu load rate LWR net destruction frac ~ LWR reactors op life mox core frac Excess Separated Pu HTGR Cycle Pu in Waste LWR MOX...reflecting the cycle used in this type of reactor. For the HTGR , the entire core consists of plutonium fuel , therefore a core fraction is not specified...cooling time Time spent fuel unloaded from HTGR reactor must cool before permanently stored 3 years Mox core fraction Fraction of

Evaluation of gas cooling for pressurized phosphoric acid fuel cell stacks

NASA Technical Reports Server (NTRS)

Farooque, M.; Skok, A. J.; Maru, H. C.; Kothmann, R. E.; Harry, R. W.

1983-01-01

Gas cooling is a more reliable, less expensive and a more simple alternative to conventional liquid cooling for heat removal from the phosphoric acid fuel cell (PAFC). The feasibility of gas cooling has already been demonstrated in atmospheric pressure stacks. This paper presents theoretical and experimental investigation of gas cooling for pressurized PAFC. Two approaches to gas cooling, Distributed Gas Cooling (DIGAS) and Separated Gas Cooling (SGC) were considered, and a theoretical comparison on the basis of cell performance indicated SGC to be superior to DIGAS. The feasibility of SGC was experimentally demonstrated by operating a 45-cell stack for 700 hours at pressure, and determining thermal response and the effect of other related parameters.

Evaluation of Gas-Cooled Pressurized Phosphoric Acid Fuel Cells for Electric Utility Power Generation

NASA Technical Reports Server (NTRS)

Faroque, M.

1983-01-01

Gas cooling is a more reliable, less expensive and a more simple alternative to conventional liquid cooling for heat removal from the phosphoric acid fuel cell (PAFC). The feasibility of gas-cooling was already demonstrated in atmospheric pressure stacks. Theoretical and experimental investigations of gas-cooling for pressurized PAFC are presented. Two approaches to gas cooling, Distributed Gas-Cooling (DIGAS) and Separated Gas-Cooling (SGC) were considered, and a theoretical comparison on the basis of cell performance indicated SGC to be superior to DIGAS. The feasibility of SGC was experimentally demonstrated by operating a 45-cell stack for 700 hours at pressure, and determining thermal response and the effect of other related parameters.

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

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

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

Cooling Characteristics of an Experimental Tail-pipe Burner with an Annular Cooling-air Passage

NASA Technical Reports Server (NTRS)

Kaufman, Harold R; Koffel, William K

1952-01-01

The effects of tail-pipe fuel-air ratio (exhaust-gas temperatures from approximately 3060 degrees to 3825 degrees R), radial distributiion of tail-pipe fuel flow, and mass flow of combustion gas and the inside wall were determined for an experimental tail-pipe burner cooled by air flowing through and insulated cooling-air to combustion gas mass flow from 0.066 to 0.192 were also determined.

Evaluation of the DRAGON code for VHTR design analysis.

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

Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

2006-01-12

This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by themore » IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.« less

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.

Thermally efficient melting and fuel reforming for glass making

DOEpatents

Chen, M.S.; Painter, C.F.; Pastore, S.P.; Roth, G.S.; Winchester, D.C.

1991-10-15

An integrated process is described for utilizing waste heat from a glass making furnace. The hot off-gas from the furnace is initially partially cooled, then fed to a reformer. In the reformer, the partially cooled off-gas is further cooled against a hydrocarbon which is thus reformed into a synthesis gas, which is then fed into the glass making furnace as a fuel. The further cooled off-gas is then recycled back to absorb the heat from the hot off-gas to perform the initial cooling. 2 figures.

Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis

DTIC Science & Technology

2016-07-31

fueled liquid rocket engine, enthalpy is removed from the combustion chamber by a regenerative cooling system comprising a series of passages through... rocket engine, enthalpy is removed from the combustion chamber by a regenerative cooling system comprising a series of passages through which fuel flows...the unprecedented correlation of comprehensive two-dimensional gas chromatographic (GC×GC) rocket fuel data with physical and thermochemical

Some advantages of methane in an aircraft gas turbine

NASA Technical Reports Server (NTRS)

Graham, R. W.; Glassman, A. J.

1980-01-01

Because liquid methane may be obtained from existing natural gas sources or produced synthetically from a range of other hydrocarbon sources (coal, biomass, shale, organic waste), it is considered as an aviation fuel in a simplified cycle analysis of the performance of a turboprop engine intended for operation at Mach 0.8 and 10,688 m altitude. Performance comparisons are given for four cases in which the turbine cooling air is either not cooled or cooled to -111, -222, and -333 K, and the advantages and problems that may be expected from direct use of the cryogenic fuel in turbine cooling are discussed. It is shown that while (1) methane combustion characteristics are appreciably different from those of Jet A fuel and will require the development of different combustor designs, and (2) the safe integration of methane cryotanks into transport aircraft structures poses a major design problem, a highly fuel-efficient turboprop engine fueled by methane appears to be feasible.

Evaluation of Foam Coolants.

DTIC Science & Technology

HYPERGOLIC ROCKET PROPELLANTS, * FOAM , FILM COOLING, FILM COOLING, LIQUID COOLING, LIQUID ROCKET FUELS, ADDITIVES, HEAT TRANSFER, COOLANTS, LIQUID PROPELLANT ROCKET ENGINES, LIQUID COOLING, CAPTIVE TESTS, FEASIBILITY STUDIES.

Thermomechanics of candidate coatings for advanced gas reactor fuels

NASA Astrophysics Data System (ADS)

Nosek, A.; Conzen, J.; Doescher, H.; Martin, C.; Blanchard, J.

2007-09-01

Candidate fuel/coating combinations for an advanced, coated-fuel particle for a gas-cooled fast reactor (GFR) have been evaluated. These all-ceramic fuel forms consist of a fuel kernel made of UC or UN, surrounded with two shells (a buffer and a coating) made of TiC, SiC, ZrC, TiN, or ZrN. These carbides and nitrides are analyzed with finite element models to determine the stresses produced in the micro fuel particles from differential thermal expansion, fission gas release, swelling, and creep during particle fabrication and reactor operation. This study will help determine the feasibility of different fuel and coating combinations and identify the critical loads. The analysis shows that differential thermal expansion of the fuel and coating dictate the amount of stress for changing temperatures (such as during fabrication), and that the coating creep is able to mitigate an otherwise overwhelming amount of stress from fuel swelling. Because fracture is a likely mode of failure, a fracture mechanics study is also included to identify the relative likelihood of catastrophic fracture of the coating and resulting gas release. Overall, the analysis predicts that UN/ZrC is the best thermomechanical fuel/coating combination for mitigating the stress within the new fuel particle, but UN/TiN and UN/ZrN could also be strong candidates if their unknown creep rates are sufficiently large.

SOFC seal and cell thermal management

DOEpatents

Potnis, Shailesh Vijay [Neenah, WI; Rehg, Timothy Joseph [Huntington Beach, CA

2011-05-17

The solid oxide fuel cell module includes a manifold, a plate, a cathode electrode, a fuel cell and an anode electrode. The manifold includes an air or oxygen inlet in communication with divergent passages above the periphery of the cell which combine to flow the air or oxygen radially or inwardly for reception in the center of the cathode flow field. The latter has interconnects providing circuitous cooling passages in a generally radial outward direction cooling the fuel cell and which interconnects are formed of different thermal conductivity materials for a preferential cooling.

An intrinsically safe facility for forefront research and training on nuclear technologies — General description of the system

NASA Astrophysics Data System (ADS)

Mansani, L.; Bruzzone, M.; Frambati, S.; Reale, M.

2014-04-01

In the framework of research on generation-IV reactors, it is very important to have infrastructures specifically dedicated to the study of fundamental parameters in dynamics and kinetics of future fast-neutron reactors. Among various options pursued by international groups, Italy focused on lead-cooled reactors, which guarantee minimal neutron slowdown and capture and efficient cooling. In this paper it is described the design of a the low-power prototype generator, LEADS, that could be used within research facilities such as the National Laboratory of Legnaro of the INFN. The LEADS has a high safety standard in order to be used as a training facility, but it has also a good flexibility so as to allow a wide range of measurements and experiments. A high safety standard is achieved by limiting the reactor power to less than few hundred kW and the neutron multiplication factor k eff to less than 0.95 (a limiting value for spent fuel pool), by using a pure-uranium fuel (no plutonium) and by using solid lead as a diffuser. The proposed core is therefore intrinsically subcritical and has to be driven by an external neutron source generated by a proton beam impinging in a target. Preliminary simulations, performed with the MCNPX code indicated, for a 0.75mA continuous proton beam current at 70MeV proton energy, a reactor power of about 190kW when using a beryllium converter. The enriched-uranium fuel elements are immersed in a solid-lead matrix and contained within a steel vessel. The system is cooled by helium gas, which is transparent to neutrons and does not undergo activation. The gas is pumped by a compressor through specific holes at the entrance of the active volume with a temperature which varies according to the operating conditions and a pressure of about 1.1MPa. The hot gas coming out of the vessel is cooled by an external helium-water heat exchanger. The beryllium converter is cooled by its dedicated helium gas cooling system. After shutdown, the decay is completely dissipated by conduction through the lead reflector and steel vessel, and then evacuated by irradiation from the vessel surface to the external ambient air.

NEUTRONIC REACTOR

DOEpatents

Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

1959-03-24

A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

Control rod system useable for fuel handling in a gas-cooled nuclear reactor

DOEpatents

Spurrier, Francis R.

1976-11-30

A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

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.

Crash-Fire Protection System for T-56 Turbopropeller Engine Using Water as Cooling and Inerting Agent

NASA Technical Reports Server (NTRS)

Busch, Arthur M.; Campbell, John A.

1959-01-01

A crash-fire protection system to suppress the ignition of crash-spilled fuel that may be ingested by a T-56 turbopropeller engine is described. This system includes means for rapidly extinguishing the combustor flame and means for cooling and inerting with water the hot engine parts likely to ignite engine-ingested fuel. Combustion-chamber flames were extinguished in 0.07 second at the engine fuel manifold. Hot engine parts were inerted and cooled by 52 pounds of water discharged at ten engine stations. Performance trials of the crash-fire prevention system were conducted by bringing the engine up to takeoff temperature, stopping the normal fuel flow to the engine, starting the water discharge, and then spraying fuel into the engine to simulate crash-ingested fuel. No fires occurred during these trials, although fuel was sprayed into the engine from 0.3 second to 15 minutes after actuating the crash-fire protection system.

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

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

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

Fuel cell system configurations

DOEpatents

Kothmann, Richard E.; Cyphers, Joseph A.

1981-01-01

Fuel cell stack configurations having elongated polygonal cross-sectional shapes and gaskets at the peripheral faces to which flow manifolds are sealingly affixed. Process channels convey a fuel and an oxidant through longer channels, and a cooling fluid is conveyed through relatively shorter cooling passages. The polygonal structure preferably includes at least two right angles, and the faces of the stack are arranged in opposite parallel pairs.

Experimental results for the rapid determination of the freezing point of fuels

NASA Technical Reports Server (NTRS)

Mathiprakasam, B.

1984-01-01

Two methods for the rapid determination of the freezing point of fuels were investigated: an optical method, which detected the change in light transmission from the disappearance of solid particles in the melted fuel; and a differential thermal analysis (DTA) method, which sensed the latent heat of fusion. A laboratory apparatus was fabricated to test the two methods. Cooling was done by thermoelectric modules using an ice-water bath as a heat sink. The DTA method was later modified to eliminate the reference fuel. The data from the sample were digitized and a point of inflection, which corresponds to the ASTM D-2386 freezing point (final melting point), was identified from the derivative. The apparatus was modifified to cool the fuel to -60 C and controls were added for maintaining constant cooling rate, rewarming rate, and hold time at minimum temperature. A parametric series of tests were run for twelve fuels with freezing points from -10 C to -50 C, varying cooling rate, rewarming rate, and hold time. Based on the results, an optimum test procedure was established. The results showed good agreement with ASTM D-2386 freezing point and differential scanning calorimetry results.

Study of active cooling for supersonic transports

NASA Technical Reports Server (NTRS)

Brewer, G. D.; Morris, R. E.

1975-01-01

The potential benefits of using the fuel heat sink of hydrogen fueled supersonic transports for cooling large portions of the aircraft wing and fuselage are examined. The heat transfer would be accomplished by using an intermediate fluid such as an ethylene glycol-water solution. Some of the advantages of the system are: (1) reduced costs by using aluminum in place of titanium, (2) reduced cabin heat loads, and (3) more favorable environmental conditions for the aircraft systems. A liquid hydrogen fueled, Mach 2.7 supersonic transport aircraft design was used for the reference uncooled vehicle. The cooled aircraft designs were analyzed to determine their heat sink capability, the extent and location of feasible cooled surfaces, and the coolant passage size and spacing.

Correlation of the Characteristics of Single-Cylinder and Flight Engines in Tests of High-Performance Fuels in an Air-Cooled Engine I : Cooling Characteristics

NASA Technical Reports Server (NTRS)

Wilson, Robert W.; Richard, Paul H.; Brown, Kenneth D.

1945-01-01

Variable charge-air flow, cooling-air pressure drop, and fuel-air ration investigations were conducted to determine the cooling characteristics of a full-scale air-cooled single cylinder on a CUE setup. The data are compared with similar data that were available for the same model multicylinder engine tested in flight in a four-engine airplane. The cylinder-head cooling correlations were the same for both the single-cylinder and the flight engine. The cooling correlations for the barrels differed slightly in that the barrel of the single-cylinder engine runs cooler than the barrel of te flight engine for the same head temperatures and engine conditions.

Fuel cell gas management system

DOEpatents

DuBose, Ronald Arthur

2000-01-11

A fuel cell gas management system including a cathode humidification system for transferring latent and sensible heat from an exhaust stream to the cathode inlet stream of the fuel cell; an anode humidity retention system for maintaining the total enthalpy of the anode stream exiting the fuel cell equal to the total enthalpy of the anode inlet stream; and a cooling water management system having segregated deionized water and cooling water loops interconnected by means of a brazed plate heat exchanger.

Effect of cooled EGR on performance and exhaust gas emissions in EFI spark ignition engine fueled by gasoline and wet methanol blends

NASA Astrophysics Data System (ADS)

Rohadi, Heru; Syaiful, Bae, Myung-Whan

2016-06-01

Fuel needs, especially the transport sector is still dominated by fossil fuels which are non-renewable. However, oil reserves are very limited. Furthermore, the hazardous components produced by internal combustion engine forces many researchers to consider with alternative fuel which is environmental friendly and renewable sources. Therefore, this study intends to investigate the impact of cooled EGR on the performance and exhaust gas emissions in the gasoline engine fueled by gasoline and wet methanol blends. The percentage of wet methanol blended with gasoline is in the range of 5 to 15% in a volume base. The experiment was performed at the variation of engine speeds from 2500 to 4000 rpm with 500 intervals. The re-circulated exhaust gasses into combustion chamber was 5%. The experiment was performed at the constant engine speed. The results show that the use of cooled EGR with wet methanol of 10% increases the brake torque up to 21.3%. The brake thermal efficiency increases approximately 39.6% using cooled EGR in the case of the engine fueled by 15% wet methanol. Brake specific fuel consumption for the engine using EGR fueled by 10% wet methanol decreases up to 23% at the engine speed of 2500 rpm. The reduction of CO, O2 and HC emissions was found, while CO2 increases.

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

NASA Astrophysics Data System (ADS)

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

1993-01-01

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

Hydrocarbon-fuel/copper combustion chamber liner compatibility, corrosion prevention, and refurbishment

NASA Technical Reports Server (NTRS)

Rosenberg, S. D.; Gage, M. L.; Homer, G. D.; Franklin, J. E.

1991-01-01

An evaluation is made of combustion product/combustion chamber compatibility in the case of a LOX/liquid hydrocarbon booster engine based on copper-alloy thrust chamber which is regeneratively cooled by the fuel. It is found that sulfur impurities in the fuel are the primary causes of copper corrosion, through formation of Cu2S; sulfur levels as low as 1 ppm can result in sufficiently severe copper corrosion to degrade cooling channel performance. This corrosion can be completely eliminated, however, through the incorporation of an electrodeposited gold coating on the copper cooling-channel walls.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Astrophysics Data System (ADS)

Emrich, William J.

2008-01-01

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

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2008-01-01

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

PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

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

Hales, J. D.; Gamble, K. A.

2015-09-01

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normalmore » operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.« less

Theoretical Estimate of Maximum Possible Nuclear Explosion

DOE R&D Accomplishments Database

Bethe, H. A.

1950-01-31

The maximum nuclear accident which could occur in a Na-cooled, Be moderated, Pu and power producing reactor is estimated theoretically. (T.R.H.) 2O82 Results of nuclear calculations for a variety of compositions of fast, heterogeneous, sodium-cooled, U-235-fueled, plutonium- and power-producing reactors are reported. Core compositions typical of plate-, pin-, or wire-type fuel elements and with uranium as metal, alloy, and oxide were considered. These compositions included atom ratios in the following range: U-23B to U-235 from 2 to 8; sodium to U-235 from 1.5 to 12; iron to U-235 from 5 to 18; and vanadium to U-235 from 11 to 33. Calculations were performed to determine the effect of lead and iron reflectors between the core and blanket. Both natural and depleted uranium were evaluated as the blanket fertile material. Reactors were compared on a basis of conversion ratio, specific power, and the product of both. The calculated results are in general agreement with the experimental results from fast reactor assemblies. An analysis of the effect of new cross-section values as they became available is included. (auth)

Study of the collector/heat pipe cooled externally configured thermionic diode

NASA Technical Reports Server (NTRS)

1973-01-01

A collector/heat pipe cooled, externally configured (heated) thermionic diode module was designed for use in a laboratory test to demonstrate the applicability of this concept as the fuel element/converter module of an in-core thermionic electric power source. During the course of the program, this module evolved from a simple experimental mock-up into an advanced unit which was more reactor prototypical. Detailed analysis of all diode components led to their engineering design, fabrication, and assembly, with the exception of the collector/heat pipe. While several designs of high power annular wicked heat pipes were fabricated and tested, each exhibited unexpected performance difficulties. It was concluded that the basic cause of these problems was the formation of crud which interfered with the liquid flow in the annular passage of the evaporator region.

Analytical and numerical study on cooling flow field designs performance of PEM fuel cell with variable heat flux

NASA Astrophysics Data System (ADS)

Afshari, Ebrahim; Ziaei-Rad, Masoud; Jahantigh, Nabi

2016-06-01

In PEM fuel cells, during electrochemical generation of electricity more than half of the chemical energy of hydrogen is converted to heat. This heat of reactions, if not exhausted properly, would impair the performance and durability of the cell. In general, large scale PEM fuel cells are cooled by liquid water that circulates through coolant flow channels formed in bipolar plates or in dedicated cooling plates. In this paper, a numerical method has been presented to study cooling and temperature distribution of a polymer membrane fuel cell stack. The heat flux on the cooling plate is variable. A three-dimensional model of fluid flow and heat transfer in cooling plates with 15 cm × 15 cm square area is considered and the performances of four different coolant flow field designs, parallel field and serpentine fields are compared in terms of maximum surface temperature, temperature uniformity and pressure drop characteristics. By comparing the results in two cases, the constant and variable heat flux, it is observed that applying constant heat flux instead of variable heat flux which is actually occurring in the fuel cells is not an accurate assumption. The numerical results indicated that the straight flow field model has temperature uniformity index and almost the same temperature difference with the serpentine models, while its pressure drop is less than all of the serpentine models. Another important advantage of this model is the much easier design and building than the spiral models.

Fuel cell design and assembly

DOEpatents

Myerhoff, Alfred

1984-01-01

The present invention is directed to a novel bipolar cooling plate, fuel cell design and method of assembly of fuel cells. The bipolar cooling plate used in the fuel cell design and method of assembly has discrete opposite edge and means carried by the plate defining a plurality of channels extending along the surface of the plate toward the opposite edges. At least one edge of the channels terminates short of the edge of the plate defining a recess for receiving a fastener.

Modeling of Fuel Film Cooling on Chamber Hot Wall

DTIC Science & Technology

2014-07-01

downstream, when the film has been depleted of its cooling and coking capacities, a second slot is needed to inject fresh cool fuel. All of these...pyrolysis and oxidation. 7. As discussed in the introductory section, sooting and coking are notoriously complex topics. Well- validated global...accurate models for soot formation and deposition. Instead, the potential impact of the coke layer is evaluated parametrically by representing the

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

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

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement.

PubMed

Škarohlíd, Jan; Ashcheulov, Petr; Škoda, Radek; Taylor, Andrew; Čtvrtlík, Radim; Tomáštík, Jan; Fendrych, František; Kopeček, Jaromír; Cháb, Vladimír; Cichoň, Stanislav; Sajdl, Petr; Macák, Jan; Xu, Peng; Partezana, Jonna M; Lorinčík, Jan; Prehradná, Jana; Steinbrück, Martin; Kratochvílová, Irena

2017-07-25

In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100-170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

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

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

CHAIN REACTING SYSTEM

DOEpatents

Fermi, E.; Leverett, M.C.

1958-06-01

A nuclear reactor of the gas-cooled, graphitemoderated type is described. In this design, graphite blocks are arranged in a substantially cylindrical lattice having vertically orienied coolant channels in which uranium fuel elements having through passages are disposed. The active lattice is contained within a hollow body. such as a steel shell, which, in turn, is surrounded by water and concrete shields. Helium is used as the primary coolant and is circulated under pressure through the coolant channels and fuel elements. The helium is then conveyed to heat exchangers, where its heat is used to produce steam for driving a prime mover, thence to filtering means where radioactive impurities are removed. From the filtering means the helium passes to a compressor and an after cooler and is ultimately returned to the reactor for recirculation. Control and safety rods are provided to stabilize or stop the reaction. A space is provided between the graphite lattice and the internal walls of the shell to allow for thermal expansion of the lattice during operation. This space is filled with a resilient packing, such as asbestos, to prevent the passage of helium.

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

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Development of a cryogenically cooled platform for the Magnetized Liner Inertial Fusion (MagLIF) Program [Development of a cryogenically-cooled platform for the Magnetized Liner Inertial Fusion (MagLIF) Concept

DOE PAGES

Awe, T. J.; Shelton, K. P.; Sefkow, A. B.; ...

2017-09-25

A cryogenically cooled hardware platform has been developed and commissioned on the Z Facility at Sandia National Laboratories in support of the Magnetized Liner Inertial Fusion (MagLIF) Program. MagLIF is a magneto-inertial fusion concept that employs a magnetically imploded metallic tube (liner) to compress and inertially confine premagnetized and preheated fusion fuel. The fuel is preheated using a ~2 kJ laser that must pass through a ~1.5-3.5-μm-thick polyimide “window” at the target’s laser entrance hole (LEH). As the terawatt-class laser interacts with the dense window, laser plasma instabilities (LPIs) can develop, which reduce the preheat energy delivered to the fuel,more » initiate fuel contamination, and degrade target performance. Cryogenically cooled targets increase the parameter space accessible to MagLIF target designs by allowing nearly 10 times thinner windows to be used for any accessible gas density. Thinner LEH windows reduce the deleterious effects of difficult to model LPIs. The Z Facility’s cryogenic infrastructure has been significantly altered to enable compatibility with the premagnetization and fuel preheat stages of MagLIF. The MagLIF cryostat brings the liquid helium coolant directly to the target via an electrically resistive conduit. This design maximizes cooling power while allowing rapid diffusion of the axial magnetic field supplied by external Helmholtz-like coils. A variety of techniques have been developed to mitigate the accumulation of ice from vacuum chamber contaminants on the cooled LEH window, as even a few hundred nanometers of ice would impact laser energy coupling to the fuel region. Here, the MagLIF cryostat has demonstrated compatibility with the premagnetization and preheat stages of MagLIF and the ability to cool targets to liquid deuterium temperatures in approximately 5 min.« less

Development of a cryogenically cooled platform for the Magnetized Liner Inertial Fusion (MagLIF) Program [Development of a cryogenically-cooled platform for the Magnetized Liner Inertial Fusion (MagLIF) Concept

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

Awe, T. J.; Shelton, K. P.; Sefkow, A. B.

A cryogenically cooled hardware platform has been developed and commissioned on the Z Facility at Sandia National Laboratories in support of the Magnetized Liner Inertial Fusion (MagLIF) Program. MagLIF is a magneto-inertial fusion concept that employs a magnetically imploded metallic tube (liner) to compress and inertially confine premagnetized and preheated fusion fuel. The fuel is preheated using a ~2 kJ laser that must pass through a ~1.5-3.5-μm-thick polyimide “window” at the target’s laser entrance hole (LEH). As the terawatt-class laser interacts with the dense window, laser plasma instabilities (LPIs) can develop, which reduce the preheat energy delivered to the fuel,more » initiate fuel contamination, and degrade target performance. Cryogenically cooled targets increase the parameter space accessible to MagLIF target designs by allowing nearly 10 times thinner windows to be used for any accessible gas density. Thinner LEH windows reduce the deleterious effects of difficult to model LPIs. The Z Facility’s cryogenic infrastructure has been significantly altered to enable compatibility with the premagnetization and fuel preheat stages of MagLIF. The MagLIF cryostat brings the liquid helium coolant directly to the target via an electrically resistive conduit. This design maximizes cooling power while allowing rapid diffusion of the axial magnetic field supplied by external Helmholtz-like coils. A variety of techniques have been developed to mitigate the accumulation of ice from vacuum chamber contaminants on the cooled LEH window, as even a few hundred nanometers of ice would impact laser energy coupling to the fuel region. Here, the MagLIF cryostat has demonstrated compatibility with the premagnetization and preheat stages of MagLIF and the ability to cool targets to liquid deuterium temperatures in approximately 5 min.« less

Unitized Regenerative Fuel Cell System Gas Dryer/Humidifier Analytical Model Development

NASA Technical Reports Server (NTRS)

Burke, Kenneth A.; Jakupca, Ian

2004-01-01

A lightweight Unitized Regenerative Fuel Cell (URFC) Energy Storage System concept is being developed at the NASA Glenn Research Center (GRC). This Unitized Regenerative Fuel Cell System (URFCS) is unique in that it uses Regenerative Gas Dryers/Humidifiers (RGD/H) that are mounted on the surface of the gas storage tanks that act as the radiators for thermal control of the Unitized Regenerative Fuel Cell System (URFCS). As the gas storage tanks cool down during URFCS charging the RGD/H dry the hydrogen and oxygen gases produced by electrolysis. As the gas storage tanks heat up during URFCS discharging, the RGD/H humidify the hydrogen and oxygen gases used by the fuel cell. An analytical model was developed to simulate the URFCS RGD/H. The model is in the form of a Microsoft (registered trademark of Microsoft Corporation) Excel worksheet that allows the investigation of the RGD/H performance. Finite Element Analysis (FEA) modeling of the RGD/H and the gas storage tank wall was also done to analyze spatial temperature distribution within the RGD/H and the localized tank wall. Test results obtained from the testing of the RGD/H in a thermal vacuum environment were used to corroborate the analyses.

Measurements of the effects of thermal contact resistance on steady state heat transfer in phosphoric-acid fuel cell stack

NASA Technical Reports Server (NTRS)

Abdul-Aziz, Ali; Alkasab, Kalil A.

1991-01-01

The influence of the thermal contact resistance on the heat transfer between the electrode plates, and the cooling system plate in a phosphoric-acid fuel-cell stack was experimentally investigated. The investigation was conducted using a set-up that simulates the operating conditions prevailing in a phosphoric acid fuel-cell stack. The fuel-cell cooling system utilized three types of coolants, water, engine oil, and air, to remove excess heat generated in the cell electrode and to maintain a reasonably uniform temperature distribution in the electrode plate. The thermal contact resistance was measured as a function of pressure at the interface between the electrode plate and the cooling system plate. The interface pressure range was from 0 kPa to 3448 kPa, while the Reynolds number for the cooling limits varied from 15 to 79 for oil, 1165 to 6165 for water, and 700 to 6864 for air. Results showed that increasing the interface pressure resulted in a higher heat transfer coefficient.

Oil cooling system for a gas turbine engine

NASA Technical Reports Server (NTRS)

Coffinberry, G. A.; Kast, H. B. (Inventor)

1977-01-01

A gas turbine engine fuel delivery and control system is provided with means to recirculate all fuel in excess fuel control requirements back to the aircraft fuel tank. This increases the fuel pump heat sink and decreases the pump temperature rise without the addition of valving other than normally employed. A fuel/oil heat exchanger and associated circuitry is provided to maintain the hot engine oil in heat exchange relationship with the cool engine fuel. Where anti-icing of the fuel filter is required, means are provided to maintain the fuel temperature entering the filter at or above a minimum level to prevent freezing thereof. In one embodiment, a divider valve is provided to take all excess fuel from either upstream or downstream of the fuel filter and route it back to the tanks, the ratio of upstream to downstream extraction being a function of fuel pump discharge pressure.

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.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, M.S.; Zawodzinski, C.

1998-08-25

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field there between for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells. 11 figs.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, Mahlon S.; Zawodzinski, Christine

2001-01-01

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, Mahlon S.; Zawodzinski, Christine

1998-01-01

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

High-pressure calorimeter chamber tests for liquid oxygen/kerosene (LOX/RP-1) rocket combustion

NASA Technical Reports Server (NTRS)

Masters, Philip A.; Armstrong, Elizabeth S.; Price, Harold G.

1988-01-01

An experimental program was conducted to investigate the rocket combustion and heat transfer characteristics of liquid oxygen/kerosene (LOX/RP-1) mixtures at high chamber pressures. Two water-cooled calorimeter chambers of different combustion lengths were tested using 37- and 61-element oxidizer-fuel-oxidizer triplet injectors. The tests were conducted at nominal chamber pressures of 4.1, 8.3, and 13.8 MPa abs (600, 1200, and 2000 psia). Heat flux Q/A data were obtained for the entire calorimeter length for oxygen/fuel mixture ratios of 1.8 to 3.3. Test data at 4.1 MPa abs compared favorably with previous test data from another source. Using an injector with a fuel-rich outer zone reduced the throat heat flux by 47 percent with only a 4.5 percent reduction in the characteristic exhaust velocity efficiency C* sub eff. The throat heat transfer coefficient was reduced approximately 40 percent because of carbon deposits on the chamber wall.

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

DOE PAGES

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; ...

2017-03-10

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

DOEpatents

Weems, S.J.

1963-09-24

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

Fuel Cell Thermal Management Through Conductive Cooling Plates

NASA Technical Reports Server (NTRS)

Colozza, Anthony J.; Burke, Kenneth A.

2008-01-01

An analysis was performed to evaluate the concept of utilizing conductive cooling plates to remove heat from a fuel cell stack, as opposed to a conventional internal cooling loop. The potential advantages of this type of cooling system are reduced stack complexity and weight and increased reliability through the reduction of the number of internal fluid seals. The conductive cooling plates would extract heat from the stack transferring it to an external coolant loop. The analysis was performed to determine the required thickness of these plates. The analysis was based on an energy balance between the thermal energy produced within the stack and the heat removal from the cooling plates. To accomplish the energy balance, the heat flow into and along the plates to the cooling fluid was modeled. Results were generated for various numbers of cells being cooled by a single cooling plate. The results provided cooling plate thickness, mass, and operating temperature of the plates. It was determined that utilizing high-conductivity pyrolitic graphite cooling plates can provide a specific cooling capacity (W/kg) equivalent to or potentially greater than a conventional internal cooling loop system.

Thermal behavior in the cracking reaction zone of scramjet cooling channels at different channel aspect ratios

NASA Astrophysics Data System (ADS)

Zhang, Silong; Feng, Yu; Jiang, Yuguang; Qin, Jiang; Bao, Wen; Han, Jiecai; Haidn, Oskar J.

2016-10-01

To study the thermal behavior in the cracking reaction zone of regeneratively cooled scramjet cooling channels at different aspect ratios, 3-D model of fuel flow in terms of the fuel's real properties and cracking reaction is built and validated through experiments. The whole cooling channel is divided into non-cracking and cracking reaction zones. Only the cracking reaction zone is studied in this article. The simulation results indicate that the fuel conversion presents a similar distribution with temperature because the fuel conversion in scramjet cooling channels is co-decided by the temperature and velocity but the temperature plays the dominate role. For the cases given in this paper, increasing the channel aspect ratio will increase the pressure drop and it is not beneficial for reducing the wall temperature because of the much severer thermal stratification, larger conversion non-uniformity, the corresponding M-shape velocity profile which will cause local heat transfer deterioration and the decreased chemical heat absorption. And the decreased chemical heat absorption caused by stronger temperature and conversion non-uniformities is bad for the utilization of chemical heat sink, chemical recuperation process and the ignition performance.

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

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

Permana, Sidik; Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132; Sekimoto, Hiroshi

2010-12-23

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period hasmore » been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.« less

Monitoring arrangement for vented nuclear fuel elements

DOEpatents

Campana, Robert J.

1981-01-01

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

Study of Compton suppression for use in spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Bender, Sarah

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

Low-thrust Isp sensitivity study

NASA Technical Reports Server (NTRS)

Schoenman, L.

1982-01-01

A comparison of the cooling requirements and attainable specific impulse performance of engines in the 445 to 4448N thrust class utilizing LOX/RP-1, LOX/Hydrogen and LOX/Methane propellants is presented. The unique design requirements for the regenerative cooling of low-thrust engines operating at high pressures (up to 6894 kPa) were explored analytically by comparing single cooling with the fuel and the oxidizer, and dual cooling with both the fuel and the oxidizer. The effects of coolant channel geometry, chamber length, and contraction ratio on the ability to provide proper cooling were evaluated, as was the resulting specific impulse. The results show that larger contraction ratios and smaller channels are highly desirable for certain propellant combinations.

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.

LCRE and SNAP 50-DR-1 programs. Engineering and progress report, April 1, 1963--June 30, 1963

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

None

BS>Declassified 6 Sep 1973. Information is presented concerning the LCRE kinetics, auxiliary systems, fuel, primary cooling system components, instrumentation, secondary cooling system, materials development, and fabrication; and SNAP-50/SPUR kinetics, fuel, primary system pump, steam generator, and materials development. (DCC)

Modeling of Fuel Film Cooling on Chamber Hot Wall

DTIC Science & Technology

2013-12-01

flow at supercritical pressure. The fuel jet and the cross-flow interact. Some part of the jet is stripped off and entrained by the hot gas...modelers. The supercritical pressure makes information on equation of state and transport properties hard to come by. The large temperature range...the modeling of hydrocarbon fuel film cooling at supercritical pressures. A relevant recent simulation study by Yang and Sun [1] used a finite-rate

NEUTRONIC REACTOR

DOEpatents

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

1958-04-15

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

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

Emrich, William J. Jr.

2008-01-21

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

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

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

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

SBIR Grant:No-Vibration Agile Cryogenic Optical Refrigerator

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

Epstein, Richard

2013-04-09

Optical refrigeration is currently the only all-solid-state cryocooling technology that has been demonstrated. Optical cryocoolers are devices that use laser light to cool small crystal or glass cooling elements. The cooling element absorbs the laser light and reradiates it at higher energy, an example of anti-Stokes fluorescence. The dif-ference between the energy of the outgoing and incoming light comes from the thermal energy of the cooling element, which in turn becomes colder. Entitled No-Vibration Agile Cryocoolers using Optical Refrigeration, this Phase I proposal directly addressed the continued development of the optical refrigerator components necessary to transition this scientific breakthrough intomore » National Nu-clear Security Administration (NNSA) sensor applications in line with the objectives of topic 50b. ThermoDynamic Films LLC (TDF), in collaboration with the University of New Mexico (UNM), cooled an optical-refrigerator cooling element comprised of an ytterbium-doped yttrium lithium fluoride (Yb:YLF) crystal from room tempera-ture to 123 K with about 2% efficiency. This is the world record in optical refrigera-tion and an important step toward revolutionizing cryogenic systems for sensor ap-plications. During this period, they also designed and analyzed the crucial elements of a prototype optical refrigerator including the thermal link that connects the cool-ing element with the load.« less

FUEL ELEMENT SUPPORT

DOEpatents

Wyman, W.L.

1961-06-27

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

Liner cooling research at NASA Lewis Research Center. [for gas turbine combustion chambers

NASA Technical Reports Server (NTRS)

Acosta, Waldo A.

1987-01-01

Described are recently completed and current advanced liner research applicable to advanced small gas turbine engines. Research relating to the evolution of fuel efficient small gas turbine engines capable of meeting future commercial and military aviation needs is currently under way at NASA Lewis Research Center. As part of this research, a reverse-flow combustor geometry was maintained while different advanced liner wall cooling techniques were investigated and compared to a baseline combustor. The performance of the combustors featuring counterflow film-cooled (CFFC) panels, transpiration cooled liner walls (TRANS), and compliant metal/ceramic (CMC) walls was obtained over a range of simulated flight conditions of a 16:1 pressure ratio gas turbine engine and fuel/air ratios up to 0.034. All the combustors featured an identical fuel injection system, identical geometric configuration outline, and similar designed internal aerothermodynamics.

Thermal breeder fuel enrichment zoning

DOEpatents

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

1992-01-01

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

Low NO[sub x], cogeneration process and system

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

Bell, R.D.

1993-07-06

A process is described for low NO[sub x] cogeneration to produce electricity and useful heat, which comprises: providing fuel and oxygen to an internal combustion engine connected to drive an electric generator, to thereby generate electricity; recovering from said engine an exhaust stream including elevated NO[sub x] levels and combined oxygen; adding to said exhaust stream sufficient fuel to create a fuel-rich mixture, the quantity of fuel being sufficient to react with the available oxygen and reduce the NO[sub x], in said exhaust stream; providing said fuel-enriched exhaust stream to a thermal reactor and reacting therein said fuel, NO[sub x]more » and available oxygen, to provide a heated oxygen-depleted stream; cooling said oxygen-depleted stream by passing same through a first heat exchanger; adding conversion oxygen to said cooled stream from said heat exchanger, and passing the cooled oxygen-augmented stream over a first catalyst bed operated at a temperature of about 750 to 1,250 F under overall reducing conditions, the quantity of conversion oxygen added being in stoichiometric excess of the amount of NO[sub x], but less than the amount of combustibles; whereby the NO[sub x] is first oxidized to NO[sub 2], and then the NO[sub 2] is reduced by the excess combustibles; cooling said stream from said first catalyst bed to a temperature of about 450 to 650 F by passing said stream through a second heat exchanger; adding air to the resulting cooled stream to produce a further cooled stream at a temperature of about 400 to 600 F, and having a stoichiometric excess of oxygen; and passing said stream having said stoichiometric excess of oxygen over an oxidizing catalyst bed at said temperature of 400 to 600 F to oxidize remaining excess combustibles, to thereby provide an effluent stream having environmentally safe characteristics.« less

Confirmation of shutdown cooling effects

NASA Astrophysics Data System (ADS)

Sato, Kotaro; Tabuchi, Masato; Sugimura, Naoki; Tatsumi, Masahiro

2015-12-01

After the Fukushima accidents, all nuclear power plants in Japan have gradually stopped their operations and have long periods of shutdown. During those periods, reactivity of fuels continues to change significantly especially for high-burnup UO2 fuels and MOX fuels due to radioactive decays. It is necessary to consider these isotopic changes precisely, to predict neutronics characteristics accurately. In this paper, shutdown cooling (SDC) effects of UO2 and MOX fuels that have unusual operation histories are confirmed by the advanced lattice code, AEGIS. The calculation results show that the effects need to be considered even after nuclear power plants come back to normal operation.

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

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

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

Low temperature chemical processing of graphite-clad nuclear fuels

DOEpatents

Pierce, Robert A.

2017-10-17

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

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

PubMed

Grambow, B; Mostafavi, M

2014-11-01

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

Impact of the water symmetry factor on humidification and cooling strategies for PEM fuel cell stacks

NASA Astrophysics Data System (ADS)

Picot, D.; Metkemeijer, R.; Bezian, J. J.; Rouveyre, L.

In this paper, experimental water and thermal balances with three proton exchange membrane fuel cells (PEMFC) are proposed. On the test facility of Ecole des Mines de Paris, three De Nora SPA fuel cell stacks have been successfully studied: An 1 kW e prototype using Nafion® 117, a 5 and a 10 kW e module using Nafion® 115. The averaged water symmetry factor determines strategies to avoid drying membrane. So, we propose analytical solutions to find compromises between humidification and cooling conditions, which determines outlet temperatures of gases. For transport applications, the space occupied by the power module must be reduced. One of the main efforts consists in decreasing the operative pressure. Thus, if adequate cooling power is applied, we show experimentally and theoretically the possibility to use De Nora PEM fuel cells with low pressure, without specific external humidification.

Integrated Cabin and Fuel Cell System Thermal Management with a Metal Hydride Heat Pump

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

Hovland, V.

2004-12-01

Integrated approaches for the heating and cooling requirements of both the fuel cell (FC) stack and cabin environment are critical to fuel cell vehicle performance in terms of stack efficiency, fuel economy, and cost. An integrated FC system and cabin thermal management system would address the cabin cooling and heating requirements, control the temperature of the stack by mitigating the waste heat, and ideally capture the waste heat and use it for useful purposes. Current work at the National Renewable Energy Laboratory (NREL) details a conceptual design of a metal hydride heat pump (MHHP) for the fuel cell system andmore » cabin thermal management.« less

Fully Premixed Low Emission, High Pressure Multi-Fuel Burner

NASA Technical Reports Server (NTRS)

Nguyen, Quang-Viet (Inventor)

2012-01-01

A low-emissions high-pressure multi-fuel burner includes a fuel inlet, for receiving a fuel, an oxidizer inlet, for receiving an oxidizer gas, an injector plate, having a plurality of nozzles that are aligned with premix face of the injector plate, the plurality of nozzles in communication with the fuel and oxidizer inlets and each nozzle providing flow for one of the fuel and the oxidizer gas and an impingement-cooled face, parallel to the premix face of the injector plate and forming a micro-premix chamber between the impingement-cooled face and the in injector face. The fuel and the oxidizer gas are mixed in the micro-premix chamber through impingement-enhanced mixing of flows of the fuel and the oxidizer gas. The burner can be used for low-emissions fuel-lean fully-premixed, or fuel-rich fully-premixed hydrogen-air combustion, or for combustion with other gases such as methane or other hydrocarbons, or even liquid fuels.

77 FR 5418 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-03

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

Uranium extraction from TRISO-coated fuel particles using supercritical CO2 containing tri-n-butyl phosphate.

PubMed

Zhu, Liyang; Duan, Wuhua; Xu, Jingming; Zhu, Yongjun

2012-11-30

High-temperature gas-cooled reactors (HTGRs) are advanced nuclear systems that will receive heavy use in the future. It is important to develop spent nuclear fuel reprocessing technologies for HTGR. A new method for recovering uranium from tristructural-isotropic (TRISO-) coated fuel particles with supercritical CO(2) containing tri-n-butyl phosphate (TBP) as a complexing agent was investigated. TRISO-coated fuel particles from HTGR fuel elements were first crushed to expose UO(2) pellet fuel kernels. The crushed TRISO-coated fuel particles were then treated under O(2) stream at 750°C, resulting in a mixture of U(3)O(8) powder and SiC shells. The conversion of U(3)O(8) into solid uranyl nitrate by its reaction with liquid N(2)O(4) in the presence of a small amount of water was carried out. Complete conversion was achieved after 60 min of reaction at 80°C, whereas the SiC shells were not converted by N(2)O(4). Uranyl nitrate in the converted mixture was extracted with supercritical CO(2) containing TBP. The cumulative extraction efficiency was above 98% after 20 min of online extraction at 50°C and 25 MPa, whereas the SiC shells were not extracted by TBP. The results suggest an attractive strategy for reprocessing spent nuclear fuel from HTGR to minimize the generation of secondary radioactive waste. Copyright © 2012 Elsevier B.V. All rights reserved.

Self-cooling mono-container fuel cell generators and power plants using an array of such generators

DOEpatents

Gillett, J.E.; Dederer, J.T.; Zafred, P.R.

1998-05-12

A mono-container fuel cell generator contains a layer of interior insulation, a layer of exterior insulation and a single housing between the insulation layers, where fuel cells, containing electrodes and electrolyte, are surrounded by the interior insulation in the interior of the generator, and the generator is capable of operating at temperatures over about 650 C, where the combination of interior and exterior insulation layers have the ability to control the temperature in the housing below the degradation temperature of the housing material. The housing can also contain integral cooling ducts, and a plurality of these generators can be positioned next to each other to provide a power block array with interior cooling. 7 figs.

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE PAGES

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.; ...

2017-04-28

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

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

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Fuel cell sub-assembly

DOEpatents

Chi, Chang V.

1983-01-01

A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

Transpiration Cooling Experiment

NASA Technical Reports Server (NTRS)

Song, Kyo D.; Ries, Heidi R.; Scotti, Stephen J.; Choi, Sang H.

1997-01-01

The transpiration cooling method was considered for a scram-jet engine to accommodate thermally the situation where a very high heat flux (200 Btu/sq. ft sec) from hydrogen fuel combustion process is imposed to the engine walls. In a scram-jet engine, a small portion of hydrogen fuel passes through the porous walls of the engine combustor to cool the engine walls and at the same time the rest passes along combustion chamber walls and is preheated. Such a regenerative system promises simultaneously cooling of engine combustor and preheating the cryogenic fuel. In the experiment, an optical heating method was used to provide a heat flux of 200 Btu/sq. ft sec to the cylindrical surface of a porous stainless steel specimen which carried helium gas. The cooling efficiencies by transpiration were studied for specimens with various porosity. The experiments of various test specimens under high heat flux have revealed a phenomenon that chokes the medium flow when passing through a porous structure. This research includes the analysis of the system and a scaling conversion study that interprets the results from helium into the case when hydrogen medium is used.

An evaporative and engine-cycle model for fuel octane sensitivity prediction

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

Moran, D.P.; Taylor, A.B.

The Motor Octane Number (MON) ranks fuels by their chemical resistance to knock. Evaporative cooling coupled with fuel chemistry determine Research Octane Number (RON) antiknock ratings. It is shown in this study that fuel Octane sensitivity (numerically RON minus MON) is liked to an important difference between the two test methods; the RON test allows each fuel`s evaporative cooling characteristics to affect gas temperature, while the MON test generally eliminates this effect by pre-evaporation. In order to establish RON test charge temperatures, a computer model of fuel evaporation was adapted to Octane Engine conditions, and simulations were compared with realmore » Octane Test Engine measurements including droplet and gas temperatures. A novel gas temperature probe yielded data that corresponded well with model predictions. Tests spanned single component fuels and blends of isomers, n-paraffins, aromatics and alcohols. Commercially available automotive and aviation gasolines were also tested. A good correlation was observed between the computer predictions and measured temperature data across the range of pure fuels and blends. A numerical method to estimate the effect of precombustion temperature differences on Octane sensitivity was developed and applied to analyze these data, and was found to predict the widely disparate sensitivities of the tested fuels with accuracy. Data are presented showing mixture temperature histories of various tested fuels, and consequent sensitivity predictions. It is concluded that a fuel`s thermal-evaporative behavior gives rise to fuel Octane sensitivity as measured by differences between the RON and MON tests. This is demonstrated by the success, over a wide range of fuels, of the sensitivity predictor method describes. Evaporative cooling, must therefore be regarded as an important parameter affecting the general road performance of automobiles.« less

Chemical Technology Division annual technical report, 1990

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

Not Available

1991-05-01

Highlights of the Chemical Technology (CMT) Division's activities during 1990 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for coal- fired magnetohydrodynamics and fluidized-bed combustion; (3) methods for recovery of energy from municipal waste and techniques for treatment of hazardous organic waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for a high-level waste repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, concentrating plutonium solids in pyrochemical residues by aqueous biphase extraction, andmore » treating natural and process waters contaminated by volatile organic compounds; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also has a program in basic chemistry research in the areas of fluid catalysis for converting small molecules to desired products; materials chemistry for superconducting oxides and associated and ordered solutions at high temperatures; interfacial processes of importance to corrosion science, high-temperature superconductivity, and catalysis; and the geochemical processes responsible for trace-element migration within the earth's crust. The Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the scientific and engineering programs at Argonne National Laboratory (ANL). 66 refs., 69 figs., 6 tabs.« less

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

Thermal expansion of the nuclear fuel-sodium reaction product Na3(U0.84(2),Na0.16(2))O4 - Structural mechanism and comparison with related sodium-metal ternary oxides

NASA Astrophysics Data System (ADS)

Illy, Marie-Claire; Smith, Anna L.; Wallez, Gilles; Raison, Philippe E.; Caciuffo, Roberto; Konings, Rudy J. M.

2017-07-01

Na3.16(2)UV,VI0.84(2)O4 is obtained from the reaction of sodium with uranium dioxide under oxygen potential conditions typical of a sodium-cooled fast nuclear reactor. In the event of a breach of the steel cladding, it would be the dominant reaction product forming at the rim of the mixed (U,Pu)O2 fuel pellets. High-temperature X-ray diffraction measurements show that a distortion of the uranium environment in Na3.16(2)UV,VI0.84(2)O4 results in a strongly anisotropic thermal expansion. A comparison with several related sodium metallates Nan-2Mn+On-1 - including Na3SbO4 and Na3TaO4, whose crystal structures are reported for the first time - has allowed us to assess the role played in the lattice expansion by the Mn+ cation radius and the Na/M ratio. On this basis, the thermomechanical behavior of the title compound is discussed, along with those of several related double oxides of sodium and actinide elements, surrogate elements, or fission products.

NUCLEAR REACTOR FUEL ELEMENT

DOEpatents

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

1961-09-01

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

Transpiration cooled throat for hydrocarbon rocket engines

NASA Technical Reports Server (NTRS)

May, Lee R.; Burkhardt, Wendel M.

1991-01-01

The objective for the Transpiration Cooled Throat for Hydrocarbon Rocket Engines Program was to characterize the use of hydrocarbon fuels as transpiration coolants for rocket nozzle throats. The hydrocarbon fuels investigated in this program were RP-1 and methane. To adequately characterize the above transpiration coolants, a program was planned which would (1) predict engine system performance and life enhancements due to transpiration cooling of the throat region using analytical models, anchored with available data; (2) a versatile transpiration cooled subscale rocket thrust chamber was designed and fabricated; (3) the subscale thrust chamber was tested over a limited range of conditions, e.g., coolant type, chamber pressure, transpiration cooled length, and coolant flow rate; and (4) detailed data analyses were conducted to determine the relationship between the key performance and life enhancement variables.

Low cost, lightweight fuel cell elements

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor)

2001-01-01

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

The relation of cool flames and auto-ignition phenomena to process safety at elevated pressure and temperature.

PubMed

Pekalski, A A; Zevenbergen, J F; Pasman, H J; Lemkowitz, S M; Dahoe, A E; Scarlett, B

2002-07-01

The cool-flame phenomenon can occur in fuel-oxygen (air) mixtures within the flammable range and outside the flammable range, at fuel-rich compositions, at temperatures below the auto-ignition temperature (AIT). It is caused by chemical reactions occurring spontaneously at relatively low temperatures and is favoured by elevated pressure. The hazards that cool flames generate are described. These vary from spoiling a product specification through contamination and explosive decomposition of condensed peroxides to the appearance of unexpected normal (hot) flame (two-stage ignition).

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

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

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

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.

Ash formation, deposition, corrosion, and erosion in conventional boilers

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

Benson, S.A.; Jones, M.L.

1995-12-01

The inorganic components (ash-forming species) associated with coals significantly affect boiler design, efficiency of operation, and lifetimes of boiler parts. During combustion in conventional pulverized fuel boilers, the inorganic components are transformed into inorganic gases, liquids, and solids. This partitioning depends upon the association of the inorganic components in the coal and combustion conditions. The inorganic components are associated as mineral grains and as organically associated elements, and these associations of inorganic components in the fuel directly influence their fate upon combustion. Combustion conditions, such as temperature and atmosphere, influence the volatility and the interaction of inorganic components during combustionmore » and gas cooling, which influences the state and size composition distribution of the particulate and condensed ash species. The intermediate species are transported with the bulk gas flow through the combustion systems, during which time the gases and entrained ash are cooled. Deposition, corrosion, and erosion occur when the ash intermediate species are transported to the heat-transfer surface, react with the surface, accumulate, sinter, and develop strength. Research over the past decade has significantly advanced understanding of ash formation, deposition, corrosion, and erosion mechanisms. Many of the advances in understanding and predicting ash-related issues can be attributed to advanced analytical methods to determine the inorganic composition of fuels and the resulting ash materials. These new analytical techniques have been the key to elucidation of the mechanisms of ash formation and deposition. This information has been used to develop algorithms and computer models to predict the effects of ash on combustion system performance.« less

Investigations into the low temperature behavior of jet fuels: Visualization, modeling, and viscosity studies

NASA Astrophysics Data System (ADS)

Atkins, Daniel L.

Aircraft operation in arctic regions or at high altitudes exposes jet fuel to temperatures below freeze point temperature specifications. Fuel constituents may solidify and remain within tanks or block fuel system components. Military and scientific requirements have been met with costly, low freeze point specialty jet fuels. Commercial airline interest in polar routes and the use of high altitude unmanned aerial vehicles (UAVs) has spurred interest in the effects of low temperatures and low-temperature additives on jet fuel. The solidification of jet fuel due to freezing is not well understood and limited visualization of fuel freezing existed prior to the research presented in this dissertation. Consequently, computational fluid dynamics (CFD) modeling that simulates jet fuel freezing and model validation were incomplete prior to the present work. The ability to simulate jet fuel freezing is a necessary tool for fuel system designers. An additional impediment to the understanding and simulation of jet fuel freezing has been the absence of published low-temperature thermo-physical properties, including viscosity, which the present work addresses. The dissertation is subdivided into three major segments covering visualization, modeling and validation, and viscosity studies. In the first segment samples of jet fuel, JPTS, kerosene, Jet A and Jet A containing additives, were cooled below their freeze point temperatures in a rectangular, optical cell. Images and temperature data recorded during the solidification process provided information on crystal habit, crystallization behavior, and the influence of the buoyancy-driven flow on freezing. N-alkane composition of the samples was determined. The Jet A sample contained the least n-alkane mass. The cooling of JPTS resulted in the least wax formation while the cooling of kerosene yielded the greatest wax formation. The JPTS and kerosene samples exhibited similar crystallization behavior and crystal habits during cooling. Low-temperature additives modified the crystal habit of the Jet A fuel. Crystal shapes and sizes were recorded for use in future computational modeling. In the second segment, a computational fluid dynamics model was developed that simulates the solidification of jet fuel due to freezing in a buoyancy-driven flow. Flow resistance caused by porous crystal structures that exist in liquid-solid regions is simulated through the use of a momentum resistance source term. (Abstract shortened by UMI.)

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE PAGES

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

Fuel-cell engine stream conditioning system

DOEpatents

DuBose, Ronald Arthur

2002-01-01

A stream conditioning system for a fuel cell gas management system or fuel cell engine. The stream conditioning system manages species potential in at least one fuel cell reactant stream. A species transfer device is located in the path of at least one reactant stream of a fuel cell's inlet or outlet, which transfer device conditions that stream to improve the efficiency of the fuel cell. The species transfer device incorporates an exchange media and a sorbent. The fuel cell gas management system can include a cathode loop with the stream conditioning system transferring latent and sensible heat from an exhaust stream to the cathode inlet stream of the fuel cell; an anode humidity retention system for maintaining the total enthalpy of the anode stream exiting the fuel cell related to the total enthalpy of the anode inlet stream; and a cooling water management system having segregated deionized water and cooling water loops interconnected by means of a brazed plate heat exchanger.

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Optimisation of air cooled, open-cathode fuel cells: Current of lowest resistance and electro-thermal performance mapping

NASA Astrophysics Data System (ADS)

Meyer, Quentin; Ronaszegi, Krisztian; Pei-June, Gan; Curnick, Oliver; Ashton, Sean; Reisch, Tobias; Adcock, Paul; Shearing, Paul R.; Brett, Daniel J. L.

2015-09-01

Selecting the ideal operating point for a fuel cell depends on the application and consequent trade-off between efficiency, power density and various operating considerations. A systematic methodology for determining the optimal operating point for fuel cells is lacking; there is also the need for a single-value metric to describe and compare fuel cell performance. This work shows how the 'current of lowest resistance' can be accurately measured using electrochemical impedance spectroscopy and used as a useful metric of fuel cell performance. This, along with other measures, is then used to generate an 'electro-thermal performance map' of fuel cell operation. A commercial air-cooled open-cathode fuel cell is used to demonstrate how the approach can be used; in this case leading to the identification of the optimum operating temperature of ∼45 °C.

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

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.

Self-cooling mono-container fuel cell generators and power plants using an array of such generators

DOEpatents

Gillett, James E.; Dederer, Jeffrey T.; Zafred, Paolo R.

1998-01-01

A mono-container fuel cell generator (10) contains a layer of interior insulation (14), a layer of exterior insulation (16) and a single housing (20) between the insulation layers, where fuel cells, containing electrodes and electrolyte, are surrounded by the interior insulation (14) in the interior (12) of the generator, and the generator is capable of operating at temperatures over about 650.degree. C., where the combination of interior and exterior insulation layers have the ability to control the temperature in the housing (20) below the degradation temperature of the housing material. The housing can also contain integral cooling ducts, and a plurality of these generators can be positioned next to each other to provide a power block array with interior cooling.

Indirect-cycle FBR cooled by supercritical steam-concept and design

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

Yoshiaki, Oka; Tatjana, Jevremovic; Sei-ichi, Koshizuka

1993-01-01

Neutronic and thermal-hydraulic design of an in direct-cycle supercritical steam-cooled fast breeder reactor (SCFBR-I) is carried out to find a way to make low-cost FBRs (Ref. 1). The advantages of supercritical steam cooling are high thermal efficiency, low pumping power, simplified system (no primary steam generators and no Loeffler boilers), and the use of experienced technology in fossil-fired power plants. The design goals are fissile fuel breeding (compound system doubling time below 30 yr), 1000-M(electric) class out-put, high fuel discharge burnup, and a long refueling period. The coolant void reactivity should be negative throughout fuel lifetime because the loss-of-coolant accidentmore » is the design-basis accident. These goals have never been satisfied simultaneously in previous SCFBRs.« less

Process for Generating Engine Fuel Consumption Map: Ricardo Cooled EGR Boost 24-bar Standard Car Engine Tier 2 Fuel

EPA Pesticide Factsheets

This document summarizes the process followed to utilize the fuel consumption map of a Ricardo modeled engine and vehicle fuel consumption data to generate a full engine fuel consumption map which can be used by EPA's ALPHA vehicle simulations.

Exploring the Use of Model-Based Systems Engineering (MBSE) to Develop Systems Architectures in Naval Ship Design

DTIC Science & Technology

2010-06-01

data such as the NSMB B-series, or be based on hydrodynamic (lifting line) predict ions. The power including still air drag and any margin that is...Provide Fuel Function 3.6 Fuel Oil System Component REQ.1.4 Fuel Efficiency Requirement 1.1 Generate Mechanical En... Function 1.1 Prime Mover Component...3.3 Provide Lubrication Function 3.7 Lube Oil System Component 3.4 Provide Cooling Water Function 3.3 Cooling System Component 3.5 Provide Combust ion

Investigation on thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing in scramjet cooling channel based on wavelet entropy method

NASA Astrophysics Data System (ADS)

Zan, Hao; Li, Haowei; Jiang, Yuguang; Wu, Meng; Zhou, Weixing; Bao, Wen

2018-06-01

As part of our efforts to find ways and means to further improve the regenerative cooling technology in scramjet, the experiments of thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing have been conducted in horizontal circular tubes at different conditions. The experimental results indicate that there is a developing process from thermo-acoustic stability to instability. In order to have a deep understanding on the developing process of thermo-acoustic instability, the method of Multi-scale Shannon Wavelet Entropy (MSWE) based on Wavelet Transform Correlation Filter (WTCF) and Multi-Scale Shannon Entropy (MSE) is adopted in this paper. The results demonstrate that the developing process of thermo-acoustic instability from noise and weak signals is well detected by MSWE method and the differences among the stability, the developing process and the instability can be identified. These properties render the method particularly powerful for warning thermo-acoustic instability of hydrocarbon fuel flowing in scramjet cooling channels. The mass flow rate and the inlet pressure will make an influence on the developing process of the thermo-acoustic instability. The investigation on thermo-acoustic instability dynamic characteristics at supercritical pressure based on wavelet entropy method offers guidance on the control of scramjet fuel supply, which can secure stable fuel flowing in regenerative cooling system.

Determination of the plutonium content in a spent fuel assembly by passive and active interrogation using a differential die-away instrument

NASA Astrophysics Data System (ADS)

Henzl, V.; Croft, S.; Richard, J.; Swinhoe, M. T.; Tobin, S. J.

2013-06-01

In this paper, we present a novel approach to estimating the total plutonium content in a spent fuel assembly (SFA) that is based on combining information from a passive measurement of the total neutron count rate (PN) of the assayed SFA and a measure of its multiplication. While PN can be measured essentially with any non-destructive assay (NDA) technique capable of neutron detection, the measure of multiplication is, in our approach, determined by means of active interrogation using an instrument based on the Differential Die-Away technique (DDA). The DDA is a NDA technique developed within the U.S. Department of Energy's Next Generation Safeguards Initiative (NGSI) project focused on the utilization of NDA techniques to determine the elemental plutonium content in commercial nuclear SFA's [1]. This approach was adopted since DDA also allows determination of other SFA characteristics, such as burnup, initial enrichment, and cooling time, and also allows for detection of certain types of diversion of nuclear material. The quantification of total plutonium is obtained using an analytical correlation function in terms of the observed PN and active multiplication. Although somewhat similar approaches relating Pu content with PN have been adopted in the past, we demonstrate by extensive simulation of the fuel irradiation and NDA process that our analytical method is independent of explicit knowledge of the initial enrichment, burnup, and an absolute value of the SFA's reactivity (i.e. multiplication factor). We show that when tested with MCNPX™ simulations comprising the 64 SFA NGSI Spent Fuel Library-1 we were able to determine elemental plutonium content, using just a few calibration parameters, with an average variation in the prediction of around 1-2% across the wide dynamic range of irradiation history parameters used, namely initial enrichment (IE=2-5%), burnup (BU=15-60 GWd/tU) and cooling time (CT=1-80 y). In this paper we describe the basic approach and the success obtained against synthetic data. We recognize that our synthetic data may not fully capture the rich behavior of actual irradiated fuel and the uncertainties of the practical measurements. However, this design study is based on a rather complete nuclide inventory and the correlations for Pu seem robust to variation of input. Thus it is concluded that the proposed method is sufficiently promising that further experimentally based work is desirable.

A simple model for closure temperature of a trace element in cooling bi-mineralic systems

NASA Astrophysics Data System (ADS)

Liang, Yan

2015-09-01

Closure temperature is defined as the lower temperature limit at which the element of interest effectively ceases diffusive exchange with its surrounding medium during cooling. Here we generalize the classic equation of Dodson (1973) for cooling mono-mineralic systems to cooling bi-mineralic aggregates by considering diffusive exchange of a trace element between the two minerals in a closed system. We present a simple analytical model that includes key parameters affecting the closure temperature of a trace element in cooling bi-mineralic systems: cooling rate, temperature-dependent diffusion coefficients for the trace element in the two minerals, temperature-dependent partition coefficient of the trace element between the two minerals, effective grain sizes of the two minerals, and volume proportions of the minerals in the system. We show that closure temperatures of a trace element in cooling bi-mineralic systems are bounded by the closure temperatures of the trace element in the two mono-mineralic systems and that our generalized model reduces to Dodson's equation when one of the mineral serves as "an effective infinite" reservoir to the other mineral. Application to closure temperatures of REE in orthopyroxene and clinopyroxene bi-mineralic systems highlights the importance of REE diffusion and partitioning in the pyroxenes as well as clinopyroxene modal abundance and grain size in the systems. Closure temperatures for REE in two-pyroxene bearing equigranular rocks are controlled primarily by diffusion in orthopyroxene unless the modal abundance of clinopyroxene is very small. This has important bearings on the interpretation of temperatures derived from the REE-in-two-pyroxene thermometer.

Cooling water distribution system

DOEpatents

Orr, Richard

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5

NASA Astrophysics Data System (ADS)

Khatry, Jivan

Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

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

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

DOE PAGES

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; ...

2015-10-26

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

NASA Technical Reports Server (NTRS)

Hsieh, T.-M.; Koenig, D. R.

1977-01-01

Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

Oxide dispersion strengthened ferritic steels: a basic research joint program in France

NASA Astrophysics Data System (ADS)

Boutard, J.-L.; Badjeck, V.; Barguet, L.; Barouh, C.; Bhattacharya, A.; Colignon, Y.; Hatzoglou, C.; Loyer-Prost, M.; Rouffié, A. L.; Sallez, N.; Salmon-Legagneur, H.; Schuler, T.

2014-12-01

AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9-14% Cr ferritic-martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical-chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α‧ unmixing.

CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

DOEpatents

Weems, S.J.

1963-09-24

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

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

Passive cooling system for top entry liquid metal cooled nuclear reactors

DOEpatents

Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

1992-01-01

A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

Implications for Metallographic Cooling Rates, Derived from Fine-Scale Analytical Traverses Across Kamacite, Taenite, and Tetrataenite in the Butler Iron Meteorite

NASA Technical Reports Server (NTRS)

Jones, J. H.; Ross, D. K.; Chabot, N. L.; Keller, L. P.

2016-01-01

The "M-shaped" Ni concentrations across Widmanstatten patterns in iron meteorites, mesosiderites, and ordinary chondrites are commonly used to calculate cooling rates. As Ni-poor kamacite exolves from Ni-rich taenite, Ni concentrations build up at the kamacite-taenite interface because of the sluggish diffusivity of Ni. Quantitative knowledge of experimentally-determined Ni diffusivities, coupled with the shape of the M-profile, have been used to allow calculation of cooling rates that pertained at low temperatures, less than or equal to 500 C. However, determining Ni metallographic cooling rates are challenging, due to the sluggish diffusivity of Ni at low temperatures. There are three potential difficulties in using Ni cooling rates at low temperatures: (i) Ni diffusivities are typically extrapolated from higher-temperature measurements; (ii) Phase changes occur at low temperatures that may be difficult to take into account; and (iii) It appears that Ge in kamacite and taenite has continued to equilibrate (or attempted to equilibrate) at temperatures below those that formed the M-shaped Ni profile. Combining Ni measurements with those of other elements has the potential to provide a way to confirm or challenge Ni-determined cooling rates, as well as provide insight into the partitioning behaviors of elements during the cooling of iron meteorites. Despite these benefits, studies that examine elemental profiles of Ni along with other elements in iron meteorites are limited, often due to the low concentration levels of the other elements and associated analytical challenges. The Butler iron meteorite provides a good opportunity to conduct a multi-element analytical study, due to the higher concentration levels of key elements in addition to Fe and Ni. In this work, we perform combined analysis for six elements in the Butler iron to determine the relative behaviors of these elements during the evolution of iron meteorites, with implications for metallographic cooling rates.

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

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

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

1961-08-01

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

Cooling assembly for fuel cells

DOEpatents

Kaufman, Arthur; Werth, John

1990-01-01

A cooling assembly for fuel cells having a simplified construction whereby coolant is efficiently circulated through a conduit arranged in serpentine fashion in a channel within a member of such assembly. The channel is adapted to cradle a flexible, chemically inert, conformable conduit capable of manipulation into a variety of cooling patterns without crimping or otherwise restricting of coolant flow. The conduit, when assembled with the member, conforms into intimate contact with the member for good thermal conductivity. The conduit is non-corrodible and can be constructed as a single, manifold-free, continuous coolant passage means having only one inlet and one outlet.

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

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

1962-08-14

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

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.

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

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

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

Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

Combustion Characteristics in a Non-Premixed Cool-Flame Regime of n-Heptane in Microgravity

NASA Technical Reports Server (NTRS)

Takahashi, Fumiaki; Katta, Viswanath R.; Hicks, Michael C.

2015-01-01

A series of distinct phenomena have recently been observed in single-fuel-droplet combustion tests performed on the International Space Station (ISS). This study attempts to simulate the observed flame behavior numerically using a gaseous n-heptane fuel source in zero gravity and a time-dependent axisymmetric (2D) code, which includes a detailed reaction mechanism (127 species and 1130 reactions), diffusive transport, and a radiation model (for CH4, CO, CO2, H2O, and soot). The calculated combustion characteristics depend strongly on the air velocity around the fuel source. In a near-quiescent air environment (< or = 2 mm/s), with a sufficiently large fuel injection velocity (1 cm/s), a growing spherical diffusion flame extinguishes at ˜1200 K due to radiative heat losses. This is typically followed by a transition to the low-temperature (cool-flame) regime with a reaction zone (at ˜700 K) in close proximity to the fuel source. The 'cool flame' regime is formed due to the negative temperature coefficient in the low-temperature chemistry. After a relatively long period (˜18 s) of the cool flame regime, a flash re-ignition occurs, associated with flame-edge propagation and subsequent extinction of the re-ignited flame. In a low-speed (˜3 mm/s) airstream (which simulates the slight droplet movement), the diffusion flame is enhanced upstream and experiences a local extinction downstream at ˜1200 K, followed by steady flame pulsations (˜0.4 Hz). At higher air velocities (4-10 mm/s), the locally extinguished flame becomes steady state. The present axisymmetric computational approach helps in revealing the non-premixed 'cool flame' structure and 2D flame-flow interactions observed in recent microgravity droplet combustion experiments.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

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

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

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

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

Hanson, A. L.; Diamond, D.

2013-10-31

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

Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

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

Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

2013-07-01

The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in themore » Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)« less

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

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

Gorgemans, J.; Mulhollem, L.; Glavin, J.

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first levelmore » of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool heat load conditions. - After 3 days, several different means are provided to continue spent fuel cooling using installed plant equipment as well as off-site equipment with built-in connections. Even for beyond design basis accidents with postulated pool damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 multiple spent fuel pool spray and fill systems provide additional lines of defense to prevent spent fuel damage. (authors)« less

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.

Cooling apparatus with a resilient heat conducting member

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

Chainer, Timothy J.; Parida, Pritish R.; Schultz, Mark D.

2016-06-14

A cooling structure including a thermally conducting central element having a channel formed therein, the channel being configured for flow of cooling fluid there through, a first pressure plate, and a first thermally conductive resilient member disposed between the thermally conducting central element and the first pressure plate, wherein the first pressure plate, the first thermally conductive resilient member, and the thermally conducting central element form a first heat transfer path.

Reactions of calcium orthosilicate and barium zirconate with oxides and sulfates of various elements

NASA Technical Reports Server (NTRS)

Zaplatynsky, I.

1979-01-01

Calcium orthosilicate and barium zirconate were evaluated as the insulation layer of thermal barrier coatings for air cooled gas turbine components. Their reactions with various oxides and sulfates were studied at 1100 C and 1300 C for times ranging up to 400 and 200 hours, respectively. These oxides and sulfates represent potential impurities or additives in gas turbine fuels and in turbine combustion air, as well as elements of potential bond coat alloys. The phase compositions of the reaction products were determined by X-ray diffraction analysis. BaZrO3 and 2CaO-SiO2 both reacted with P2O5, V2O5, Cr2O3, Al2O3, and SiO2. In addition, 2CaO-SiO2 reacted with Na2O, BaO, MgO, and CoO and BaZrO3 reacted with Fe2O3.

Pm-1 Reactor Core Final Design Report

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

Bagley, R. O.; Cox, F. H.; Carnasale, A.

1962-01-01

The PM-1 water cooled and moderated core contains 741 highly enriched stainless steel cermet tubular fuel elements and 90 lumped B stainless steel burnable poison elements, and it is controlled by 6 Y-shaped europium titanate movable control rods. The core has a lifetime of 1.95 years when operated at its design power level of 9.37 mw of thermal energy. The control of the core is designed so that there is a positive shutdown margin at all times with either one rod stuck completely out or the core or with two rods stuck in the operating condition. The core power ismore » removed by 2125 gpm of pressurized water at an average temperature of 463 deg F and pressure of 1300 psia. In reactors of this type, the core is stable with a negative temperature coefficient of approximately 2.5 x 10/sup -4/ DELTA K/K/ deg F.« less

Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

NASA Astrophysics Data System (ADS)

Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

2012-06-01

Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

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.

Advanced diesel electronic fuel injection and turbocharging

NASA Astrophysics Data System (ADS)

Beck, N. J.; Barkhimer, R. L.; Steinmeyer, D. C.; Kelly, J. E.

1993-12-01

The program investigated advanced diesel air charging and fuel injection systems to improve specific power, fuel economy, noise, exhaust emissions, and cold startability. The techniques explored included variable fuel injection rate shaping, variable injection timing, full-authority electronic engine control, turbo-compound cooling, regenerative air circulation as a cold start aid, and variable geometry turbocharging. A Servojet electronic fuel injection system was designed and manufactured for the Cummins VTA-903 engine. A special Servojet twin turbocharger exhaust system was also installed. A series of high speed combustion flame photos was taken using the single cylinder optical engine at Michigan Technological University. Various fuel injection rate shapes and nozzle configurations were evaluated. Single-cylinder bench tests were performed to evaluate regenerative inlet air heating techniques as an aid to cold starting. An exhaust-driven axial cooling air fan was manufactured and tested on the VTA-903 engine.

Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

DOEpatents

Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

1985-01-01

A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

Effect on combined cycle efficiency of stack gas temperature constraints to avoid acid corrosion

NASA Technical Reports Server (NTRS)

Nainiger, J. J.

1980-01-01

To avoid condensation of sulfuric acid in the gas turbine exhaust when burning fuel oils contaning sulfur, the exhaust stack temperature and cold-end heat exchanger surfaces must be kept above the condensation temperature. Raising the exhaust stack temperature, however, results in lower combined cycle efficiency compared to that achievable by a combined cycle burning a sulfur-free fuel. The maximum difference in efficiency between the use of sulfur-free and fuels containing 0.8 percent sulfur is found to be less than one percentage point. The effect of using a ceramic thermal barrier coating (TBC) and a fuel containing sulfur is also evaluated. The combined-cycle efficiency gain using a TBC with a fuel containing sulfur compared to a sulfur-free fuel without TBC is 0.6 to 1.0 percentage points with air-cooled gas turbines and 1.6 to 1.8 percentage points with water-cooled gas turbines.

Nuclear reactor

DOEpatents

Thomson, Wallace B.

2004-03-16

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

Power generating system and method utilizing hydropyrolysis

DOEpatents

Tolman, R.

1986-12-30

A vapor transmission cycle is described which burns a slurry of coal and water with some of the air from the gas turbine compressor, cools and cleans the resulting low-Btu fuel gas, burns the clean fuel gas with the remaining air from the compressor, and extracts the available energy in the gas turbine. The cycle lends itself to combined-cycle cogeneration for the production of steam, absorption cooling, and electric power.

Fail-safe system for activity cooled supersonic and hypersonic aircraft. [using liquid hydrogen fuel

NASA Technical Reports Server (NTRS)

Jones, R. A.; Braswell, D. O.; Richie, C. B.

1975-01-01

A fail-safe-system concept was studied as an alternative to a redundant active cooling system for supersonic and hypersonic aircraft which use the heat sink of liquid-hydrogen fuel for cooling the aircraft structure. This concept consists of an abort maneuver by the aircraft and a passive thermal protection system (TPS) for the aircraft skin. The abort manuever provides a low-heat-load descent from normal cruise speed to a lower speed at which cooling is unnecessary, and the passive TPS allows the aircraft skin to absorb the abort heat load without exceeding critical skin temperature. On the basis of results obtained, it appears that this fail-safe-system concept warrants further consideration, inasmuch as a fail-safe system could possibly replace a redundant active cooling system with no increase in weight and would offer other potential advantages.

Triaxial Swirl Injector Element for Liquid-Fueled Engines

NASA Technical Reports Server (NTRS)

Muss, Jeff

2010-01-01

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

Feasibility study of a fission-suppressed Tokamak fusion breeder

NASA Astrophysics Data System (ADS)

Moir, R. W.; Lee, J. D.; Neef, W. S., Jr.; Berwald, D. H.; Garner, J. K.; Whitley, R. H.; Ghoniem, N.; Wong, C. P. C.; Maya, I.; Schultz, K. R.

1984-12-01

The preliminary conceptual design of a tokama fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m(2) and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 plus or minus 30% per fusion reaction. This results in the production of 4900 kg of (223)U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW sub e LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U308 depending on government financing or utility financing assumptions. Additional topics discussed include the Tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

Qualitative Flow Visualization of a 110-N Hydrogen/Oxygen Laboratory Model Thruster

NASA Technical Reports Server (NTRS)

deGroot, Wim A.; McGuire, Thomas J.; Schneider, Steven J.

1997-01-01

The flow field inside a 110 N gaseous hydrogen/oxygen thruster was investigated using an optically accessible, two-dimensional laboratory test model installed in a high altitude chamber. The injector for this study produced an oxidizer-rich core flow, which was designed to fully mix and react inside a fuel-film sleeve insert before emerging into the main chamber section, where a substantial fuel film cooling layer was added to protect the chamber wall. Techniques used to investigate the flow consisted of spontaneous Raman spectra measurements, visible emission imaging, ultraviolet hydroxyl spectroscopy, and high speed schlieren imaging. Experimental results indicate that the oxygen rich core flow continued to react while emerging from the fuel-film sleeve, suggesting incomplete mixing of the hydrogen in the oxygen rich core flow. Experiments also showed that the fuel film cooling protective layer retained its integrity throughout the straight section of the combustion chamber. In the converging portion of the chamber, however, a turbulent reaction zone near the wall destroyed the integrity of the film layer, a result which implies that a lower contraction angle may improve the fuel film cooling in the converging section and extend the hardware lifetime.

Foam shell cryogenic ICF target

DOEpatents

Darling, Dale H.

1987-01-01

A uniform cryogenic layer of DT fuel is maintained in a fusion target having a low density, small pore size, low Z rigid foam shell saturated with liquid DT fuel. Capillary action prevents gravitational slumping of the fuel layer. The saturated shell may be cooled to produce a solid fuel layer.

DOE Hydrogen & Fuel Cell Overview

DTIC Science & Technology

2011-01-13

Overview of Combined Heat+Power PowerElectricity Natural Gas Heat + Cooling Natural Gas or Biogas ...Fuel Cell Technologies Program eere.energy.gov Source: US DOE 10/2010 Biogas Benefits: Preliminary Analysis Stationary fuel...with the national grid. Source: US DOE 1/2011 6 | Fuel Cell Technologies Program eere.energy.gov Biogas Resource Example

Laser apparatus

DOEpatents

Lewis, Owen; Stogran, Edmund M.

1980-01-01

Laser apparatus is described wherein an active laser element, such as the disc of a face-pumped laser, is mounted in a housing such that the weight of the element is supported by glass spheres which fill a chamber defined in the housing between the walls of the housing and the edges of the laser element. The uniform support provided by the spheres enable the chamber and the pump side of the laser element to be sealed without affecting the alignment or other optical properties of the laser element. Cooling fluid may be circulated through the sealed region by way of the interstices between the spheres. The spheres, and if desired also the cooling fluid may contain material which absorbs radiation at the wavelength of parasitic emissions from the laser element. These parasitic emissions enter the spheres through the interface along the edge surface of the laser element and it is desirable that the index of refraction of the spheres and cooling fluid be near the index of refraction of the laser element. Thus support, cooling, and parasitic suppression functions are all accomplished through the use of the arrangement.

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.

Chemical Technology Division, Annual technical report, 1991

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

Not Available

1992-03-01

Highlights of the Chemical Technology (CMT) Division's activities during 1991 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for fluidized-bed combustion and coal-fired magnetohydrodynamics; (3) methods for treatment of hazardous and mixed hazardous/radioactive waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for an unsaturated repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removalmore » of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also conducts basic research in catalytic chemistry associated with molecular energy resources; chemistry of superconducting oxides and other materials of interest with technological application; interfacial processes of importance to corrosion science, catalysis, and high-temperature superconductivity; and the geochemical processes involved in water-rock interactions occurring in active hydrothermal systems. In addition, the Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the technical programs at Argonne National Laboratory (ANL).« less

Chemical Technology Division, Annual technical report, 1991

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

Not Available

1992-03-01

Highlights of the Chemical Technology (CMT) Division`s activities during 1991 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for fluidized-bed combustion and coal-fired magnetohydrodynamics; (3) methods for treatment of hazardous and mixed hazardous/radioactive waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for an unsaturated repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removalmore » of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also conducts basic research in catalytic chemistry associated with molecular energy resources; chemistry of superconducting oxides and other materials of interest with technological application; interfacial processes of importance to corrosion science, catalysis, and high-temperature superconductivity; and the geochemical processes involved in water-rock interactions occurring in active hydrothermal systems. In addition, the Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the technical programs at Argonne National Laboratory (ANL).« less

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium-Plutonium Carbide based fuel and 300 MWt Power

NASA Astrophysics Data System (ADS)

Clief Pattipawaej, Sandro; Su'ud, Zaki

2017-01-01

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

Water cooling system for an air-breathing hypersonic test vehicle

NASA Technical Reports Server (NTRS)

Petley, Dennis H.; Dziedzic, William M.

1993-01-01

This study provides concepts for hypersonic experimental scramjet test vehicles which have low cost and low risk. Cryogenic hydrogen is used as the fuel and coolant. Secondary water cooling systems were designed. Three concepts are shown: an all hydrogen cooling system, a secondary open loop water cooled system, and a secondary closed loop water cooled system. The open loop concept uses high pressure helium (15,000 psi) to drive water through the cooling system while maintaining the pressure in the water tank. The water flows through the turbine side of the turbopump to pump hydrogen fuel. The water is then allowed to vent. In the closed loop concept high pressure, room temperature, compressed liquid water is circulated. In flight water pressure is limited to 6000 psi by venting some of the water. Water is circulated through cooling channels via an ejector which uses high pressure gas to drive a water jet. The cooling systems are presented along with finite difference steady-state and transient analysis results. The results from this study indicate that water used as a secondary coolant can be designed to increase experimental test time, produce minimum venting of fluid and reduce overall development cost.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

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

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

Thermal analysis of regenerative-cooled pylon in multi-mode rocket based combined cycle engine

NASA Astrophysics Data System (ADS)

Yan, Dekun; He, Guoqiang; Li, Wenqiang; Zhang, Duo; Qin, Fei

2018-07-01

Combining pylon injector with rocket is an effective method to achieve efficient mixing and combustion in the RBCC engine. This study designs a fuel pylon with active cooling structure, and numerically investigates the coupled heat transfer between active cooling process in the pylon and combustion in the combustor in different modes. Effect of the chemical reaction of the fuel on the flow, heat transfer and physical characteristics is also discussed. The numerical results present a good agreement with the experimental data. Results indicate that drastic supplementary combustion caused by rocket gas and secondary combustion caused by the fuel injection from the pylon result in severe thermal load on the pylon. Although regenerative cooling without cracking can reduce pylon's temperature below the allowable limit, a high-temperature area appears in the middle and nail section of the pylon due to the coolant's insufficient convective heat transfer coefficient. Comparatively, endothermic cracking can provide extra chemical heat sink for the coolant and low velocity contributes to prolong the reaction time to increase the heat absorption from chemical reaction, which further lowers and unifies the pylon surface temperature.

Sectoral combustor for burning low-BTU fuel gas

DOEpatents

Vogt, Robert L.

1980-01-01

A high-temperature combustor for burning low-BTU coal gas in a gas turbine is disclosed. The combustor includes several separately removable combustion chambers each having an annular sectoral cross section and a double-walled construction permitting separation of stresses due to pressure forces and stresses due to thermal effects. Arrangements are described for air-cooling each combustion chamber using countercurrent convective cooling flow between an outer shell wall and an inner liner wall and using film cooling flow through liner panel grooves and along the inner liner wall surface, and for admitting all coolant flow to the gas path within the inner liner wall. Also described are systems for supplying coal gas, combustion air, and dilution air to the combustion zone, and a liquid fuel nozzle for use during low-load operation. The disclosed combustor is fully air-cooled, requires no transition section to interface with a turbine nozzle, and is operable at firing temperatures of up to 3000.degree. F. or within approximately 300.degree. F. of the adiabatic stoichiometric limit of the coal gas used as fuel.

Temporal Evolution of Water Use for Thermoelectric Generation

NASA Astrophysics Data System (ADS)

Reedy, R. C.; Scanlon, B. R.

2013-12-01

The long lifespan of power plants (30 - 50 yr) results in the current power plant fleet representing a legacy of past variations in fuel availability and costs, water availability and water rights, and advances in technologies, such as combined cycle plants, which impact trends in water consumption. The objective of this study was to reconstruct past water consumption and withdrawal of thermoelectric generation based on data on controls, including fuel types, generator technologies, and cooling systems, using Texas as a case study and comparing with the US. Fuel sources in Texas varied over time, from predominantly natural gas in the 1960s and early 1970s to coal and nuclear sources following the 1973 oil embargo and more recently to large increases in natural gas generation (85% increase 1998 - 2004) in response to hydraulic fracturing and low natural gas prices. The dominant generator technology in Texas was steam turbines until the early 1990s; however, combined cycle plants markedly increased in the late 1990s (400% increase 1998 - 2004). Proliferation of cooling ponds in Texas, mostly in the 1970s and 1980s (340% increase) reflects availability of large quantities of unappropriated surface water and increases in water rights permitting during this time and lower cost and higher cooling efficiency of ponds relative to wet cooling towers. Water consumption for thermoelectricity in Texas in 2010 totaled ~0.53 km3 (0.43 million acre feet, maf), accounting for ~4% of total state water consumption. High water withdrawals (32.3 km3, 26.2 maf) mostly reflect circulation between cooling ponds and power plants. About a third of the water withdrawals is not required for cooling and reflects circulation by idling plants being used as peaking plants. Controls on water consumption include (1) generator technology/thermal efficiency and (2) cooling system resulting in statewide consumption for natural gas combined cycle generators with mostly cooling towers being 60% lower than that of traditional coal, nuclear, or natural gas steam turbine generators with mostly cooling ponds. The primary control on water withdrawals is cooling system, with ~ two orders of magnitude lower withdrawals for cooling towers relative to once-through ponds statewide. Increases in natural gas combined cycle plants with cooling towers in response to high production of low-cost natural gas has greatly reduced water demand for thermoelectric cooling since 2000. A similar approach will be applied to thermoelectric generation throughout the US using information on fuel sources, generator technologies and cooling systems to better understand current water use for thermoelectric generation based on the legacy of past drivers and long lifespans of power plants. Understanding the historical evolution of water needs for thermoelectricity should allow us to better project future water needs.

Effect of Water-Alcohol Injection and Maximum Economy Spark Advance on Knock-Limited Performance and Fuel Economy of a Large Air-Cooled Cylinder

NASA Technical Reports Server (NTRS)

Heinicke, Orville H.; Vandeman, Jack E.

1945-01-01

An investigation was conducted to determine the effect of a coolant solution of 25 percent ethyl alcohol, 25 percent methyl alcohol, and 50 percent water by volume and maximum-economy spark advance on knock-limited performance and fuel economy of a large air-cooled cylinder. The knock-limited performance of the cylinder at engine speeds of 2100 and 2500 rpm was determined for coolant-fuel ratios of 0.0, 0.2, and 0.4. The effect of water-alcohol injection on fuel economy was determined in constant charge-air flow tests. The tests were conducted at a spark advance of 20 deg B.T.C. and maximum-economy spark advance.

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

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

The induction of water to the inlet air as a means of internal cooling in aircraft-engine cylinders

NASA Technical Reports Server (NTRS)

Rothrock, Addison M; Krsek, Alois, Jr; Jones, Anthony W

1943-01-01

Report presents the results of investigations conducted on a full-scale air-cooled aircraft-engine cylinder of 202-cubic inch displacement to determine the effects of internal cooling by water induction on the maximum permissible power and output of an internal-combustion engine. For a range of fuel-air and water-fuel ratios, the engine inlet pressure was increased until knock was detected aurally, the power was then decreased 7 percent holding the ratios constant. The data indicated that water was a very effective internal coolant, permitting large increases in engine power as limited by either knock or by cylinder temperatures.

Experimental and simulation study of a Gaseous oxygen/Gaseous hydrogen vortex cooling thrust chamber

NASA Astrophysics Data System (ADS)

Yu, Nanjia; Zhao, Bo; Li, Gongnan; Wang, Jue

2016-01-01

In this paper, RNG k-ε turbulence model and PDF non-premixed combustion model are used to simulate the influence of the diameter of the ring of hydrogen injectors and oxidizer-to-fuel ratio on the specific impulse of the vortex cooling thrust chamber. The simulation results and the experimental tests of a 2000 N Gaseous oxygen/Gaseous hydrogen vortex cooling thrust chamber reveal that the efficiency of the specific impulse improves significantly with increasing of the diameter of the ring of hydrogen injectors. Moreover, the optimum efficiency of the specific impulse is obtained when the oxidizer-to-fuel ratio is near the stoichiometric ratio.

A comparison of fuel savings in the residential and commercial sectors generated by the installation of solar heating and cooling systems under three tax credit scenarios

NASA Astrophysics Data System (ADS)

Moden, R.

An analysis of expected energy savings between 1977 and 1980 under three different solar tax credit scenarios is presented. The results were obtained through the solar heating and cooling of buildings (SHACOB) commercialization model. This simulation provides projected savings of conventional fuels through the installation of solar heating and cooling systems on buildings in the residential and commercial sectors. The three scenarios analyzed considered the tax credits contained in the Windfall Profits Tax of April 1980, the National Tax Act of November 1978, and a case where no tax credit is in effect.

Fuel pins with both target and fuel pellets in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

Fuel pumping system and method

DOEpatents

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

2006-12-19

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

Fuel Pumping System And Method

DOEpatents

Shafer, Scott F.; Wang, Lifeng

2005-12-13

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

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

Evaporation Loss of Light Elements as a Function of Cooling Rate: Logarithmic Law

NASA Technical Reports Server (NTRS)

Xiong, Yong-Liang; Hewins, Roger H.

2003-01-01

Knowledge about the evaporation loss of light elements is important to our understanding of chondrule formation processes. The evaporative loss of light elements (such as B and Li) as a function of cooling rate is of special interest because recent investigations of the distribution of Li, Be and B in meteoritic chondrules have revealed that Li varies by 25 times, and B and Be varies by about 10 times. Therefore, if we can extrapolate and interpolate with confidence the evaporation loss of B and Li (and other light elements such as K, Na) at a wide range of cooling rates of interest based upon limited experimental data, we would be able to assess the full range of scenarios relating to chondrule formation processes. Here, we propose that evaporation loss of light elements as a function of cooling rate should obey the logarithmic law.

Means for supporting fuel elements in a nuclear reactor

DOEpatents

Andrews, Harry N.; Keller, Herbert W.

1980-01-01

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

Fuel assembly for nuclear reactors

DOEpatents

Creagan, Robert J.; Frisch, Erling

1977-01-01

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

49 CFR 579.21 - Reporting requirements for manufacturers of 5,000 or more light vehicles annually.

Code of Federal Regulations, 2014 CFR

2014-10-01

... (compressed natural gas), CIF (compression ignition fuel), EBP (electric battery power), FCP (fuel-cell power... (electric battery power), FCP (fuel-cell power), HEV (hybrid electric vehicle), HCP (hydrogen combustion... and engine cooling system, 07 fuel system, 10 power train, 11 electrical system, 12 exterior lighting...

Gas turbine premixer with internal cooling

DOEpatents

York, William David; Johnson, Thomas Edward; Lacy, Benjamin Paul; Stevenson, Christian Xavier

2012-12-18

A system that includes a turbine fuel nozzle comprising an air-fuel premixer. The air-fuel premixed includes a swirl vane configured to swirl fuel and air in a downstream direction, wherein the swirl vane comprises an internal coolant path from a downstream end portion in an upstream direction through a substantial length of the swirl vane.

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

NASA Astrophysics Data System (ADS)

Mella, R.; Wenman, M. R.

2013-06-01

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

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

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

Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

NASA Astrophysics Data System (ADS)

Fratoni, Massimiliano

This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

NASA Glenn Research Center Program in High Power Density Motors for Aeropropulsion

NASA Technical Reports Server (NTRS)

Brown, Gerald V.; Kascak, Albert F.; Ebihara, Ben; Johnson, Dexter; Choi, Benjamin; Siebert, Mark; Buccieri, Carl

2005-01-01

Electric drive of transport-sized aircraft propulsors, with electric power generated by fuel cells or turbo-generators, will require electric motors with much higher power density than conventional room-temperature machines. Cryogenic cooling of the motor windings by the liquid hydrogen fuel offers a possible solution, enabling motors with higher power density than turbine engines. Some context on weights of various systems, which is required to assess the problem, is presented. This context includes a survey of turbine engine weights over a considerable size range, a correlation of gear box weights and some examples of conventional and advanced electric motor weights. The NASA Glenn Research Center program for high power density motors is outlined and some technical results to date are presented. These results include current densities of 5,000 A per square centimeter current density achieved in cryogenic coils, finite element predictions compared to measurements of torque production in a switched reluctance motor, and initial tests of a cryogenic switched reluctance motor.

NUCLEAR REACTOR CORE DESIGN

DOEpatents

Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

1960-03-22

An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

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

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

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

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

Variants of closing the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

2015-12-01

Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

Conversion of a micro, glow-ignition, two-stroke engine from nitromethane-methanol blend fuel to military jet propellant (JP-8)

NASA Astrophysics Data System (ADS)

Wiegand, Andrew L.

The goal of the thesis "Conversion of a Micro, Glow-Ignition, Two-Stroke Engine from Nitromethane-Methanol Blend Fuel to Military Jet Propellant (JP-8)" was to demonstrate the ability to operate a small engine on JP-8 and was completed in two phases. The first phase included choosing, developing a test stand for, and baseline testing a nitromethane-methanol-fueled engine. The chosen engine was an 11.5 cc, glow-ignition, two-stroke engine designed for remote-controlled helicopters. A micro engine test stand was developed to load and motor the engine. Instrumentation specific to the low flow rates and high speeds of the micro engine was developed and used to document engine behavior. The second phase included converting the engine to operate on JP-8, completing JP-8-fueled steady-state testing, and comparing the performance of the JP-8-fueled engine to the nitromethane-methanol-fueled engine. The conversion was accomplished through a novel crankcase heating method; by heating the crankcase for an extended period of time, a flammable fuel-air mixture was generated in the crankcase scavenged engine, which greatly improved starting times. To aid in starting and steady-state operation, yttrium-zirconia impregnated resin (i.e. ceramic coating) was applied to the combustion surfaces. This also improved the starting times of the JP-8-fueled engine and ultimately allowed for a 34-second starting time. Finally, the steady-state data from both the nitromethane-methanol and JP-8-fueled micro engine were compared. The JP-8-fueled engine showed signs of increased engine friction while having higher indicated fuel conversion efficiency and a higher overall system efficiency. The minimal ability of JP-8 to cool the engine via evaporative effects, however, created the necessity of increased cooling air flow. The conclusion reached was that JP-8-fueled micro engines could be viable in application, but not without additional research being conducted on combustion phenomenon and cooling requirements.

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

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

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

2008-07-15

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

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

DTIC Science & Technology

1990-06-01

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

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

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

A review of high-temperature polymer electrolyte membrane fuel-cell (HT-PEMFC)-based auxiliary power units for diesel-powered road vehicles

NASA Astrophysics Data System (ADS)

Liu, Yongfeng; Lehnert, Werner; Janßen, Holger; Samsun, Remzi Can; Stolten, Detlef

2016-04-01

This paper presents an extensive review of research on the development of auxiliary power units with enhanced reformate tolerance for high temperature polymer electrolyte membrane fuel cells (HT-PEMFCs). Developments in diesel reforming for fuel cells as auxiliary power units (APUs), single fuel cells and stacks and systems are outlined in detail and key findings are presented. Summaries of HT-PEMFC APU applications and start-up times for HT-PEMFC systems are then given. A summary of cooling HT-PEMFC stacks using a classic schematic diagram of a 24-cell HT-PEMFC stack, with a cooling plate for every third cell, is also presented as part of a stack analysis. Finally, a summary of CO tolerances for fuel cells is given, along with the effects of different CO volume fractions on polarization curves, the fraction of CO coverage, hydrogen coverage, anode overpotential and cell potential.

Gas-Generator Augmented Expander Cycle Rocket Engine

NASA Technical Reports Server (NTRS)

Greene, William D. (Inventor)

2011-01-01

An augmented expander cycle rocket engine includes first and second turbopumps for respectively pumping fuel and oxidizer. A gas-generator receives a first portion of fuel output from the first turbopump and a first portion of oxidizer output from the second turbopump to ignite and discharge heated gas. A heat exchanger close-coupled to the gas-generator receives in a first conduit the discharged heated gas, and transfers heat to an adjacent second conduit carrying fuel exiting the cooling passages of a primary combustion chamber. Heat is transferred to the fuel passing through the cooling passages. The heated fuel enters the second conduit of the heat exchanger to absorb more heat from the first conduit, and then flows to drive a turbine of one or both of the turbopumps. The arrangement prevents the turbopumps exposure to combusted gas that could freeze in the turbomachinery and cause catastrophic failure upon attempted engine restart.

Summary report on effects at temperature, humidity, and fuel-air ratio on two air-cooled light aircraft engines

NASA Technical Reports Server (NTRS)

Kempke, E. E., Jr.

1976-01-01

Five different engine models were tested to experimentally characterize emissions and to determine the effects of variation in fuel-air ratio and spark timing on emissions levels and other operating characteristics such as cooling, misfiring, roughness, power acceleration, etc. The results are given of two NASA reports covering the Avco Lycoming 0-320-D engine testing and the recently obtained results on the Teledyne Continental TSIO-360-C engine.

78 FR 17591 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-22

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

Operating Temperatures of a Sodium-Cooled Exhaust Valve as Measured by a Thermocouple

NASA Technical Reports Server (NTRS)

Sanders, J. C.; Wilsted, H. D.; Mulcahy, B. A.

1943-01-01

A thermocouple was installed in the crown of a sodium-cooled exhaust valve. The valve was then tested in an air-cooled engine cylinder and valve temperatures under various engine operating conditions were determined. A temperature of 1337 F was observed at a fuel-air ratio of 0.064, a brake mean effective pressure of 179 pounds per square inch, and an engine speed of 2000 rpm. Fuel-air ratio was found to have a large influence on valve temperature, but cooling-air pressure and variation in spark advance had little effect. An increase in engine power by change of speed or mean effective pressure increased the valve temperature. It was found that the temperature of the rear spark-plug bushing was not a satisfactory indication of the temperature of the exhaust valve.

Iridium-Coated Rhenium Radiation-Cooled Rockets

NASA Technical Reports Server (NTRS)

Reed, Brian D.; Biaglow, James A.; Schneider, Steven J.

1997-01-01

Radiation-cooled rockets are used for a range of low-thrust propulsion functions, including apogee insertion, attitude control, and repositioning of satellites, reaction control of launch vehicles, and primary propulsion for planetary space- craft. The key to high performance and long lifetimes for radiation-cooled rockets is the chamber temperature capability. The material system that is currently used for radiation-cooled rockets, a niobium alloy (C103) with a fused silica coating, has a maximum operating temperature of 1370 C. Temperature limitations of C103 rockets force the use of fuel film cooling, which degrades rocket performance and, in some cases, imposes a plume contamination issue from unburned fuel. A material system composed of a rhenium (Re) substrate and an iridium (Ir) coating has demonstrated operation at high temperatures (2200 C) and for long lifetimes (hours). The added thermal margin afforded by iridium-coated rhenium (Ir/Re) allows reduction or elimination of fuel film cooling. This, in turn, leads to higher performance and cleaner spacecraft environments. There are ongoing government- and industry-sponsored efforts to develop flight Ir/ Re engines, with the primary focus on 440-N, apogee insertion engines. Complementing these Ir/Re engine development efforts is a program to address specific concerns and fundamental characterization of the Ir/Re material system, including (1) development of Ir/Re rocket fabrication methods, (2) establishment of critical Re mechanical properly data, (3) development of reliable joining methods, and (4) characterization of Ir/Re life-limiting mechanisms.

Multi-Physics Simulation of TREAT Kinetics using MAMMOTH

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

DeHart, Mark; Gleicher, Frederick; Ortensi, Javier

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific fuels transient tests range from simple temperature transients to full fuel melt accidents. The current TREAT core is driven by highly enriched uranium (HEU) dispersed in amore » graphite matrix (1:10000 U-235/C atom ratio). At the center of the core, fuel is removed allowing for the insertion of an experimental test vehicle. TREAT’s design provides experimental flexibility and inherent safety during neutron pulsing. This safety stems from the graphite in the driver fuel having a strong negative temperature coefficient of reactivity resulting from a thermal Maxwellian shift with increased leakage, as well as graphite acting as a temperature sink. Air cooling is available, but is generally used post-transient for heat removal. DOE and INL have expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility, with an emphasis on effective and safe operation while minimizing experimental time and cost. At INL, the Multi-physics Object Oriented Simulation Environment (MOOSE) has been selected as the model development framework for this work. This paper describes the results of preliminary simulations of a TREAT fuel element under transient conditions using the MOOSE-based MAMMOTH reactor physics tool.« less

Standardized verification of fuel cycle modeling

DOE PAGES

Feng, B.; Dixon, B.; Sunny, E.; ...

2016-04-05

A nuclear fuel cycle systems modeling and code-to-code comparison effort was coordinated across multiple national laboratories to verify the tools needed to perform fuel cycle analyses of the transition from a once-through nuclear fuel cycle to a sustainable potential future fuel cycle. For this verification study, a simplified example transition scenario was developed to serve as a test case for the four systems codes involved (DYMOND, VISION, ORION, and MARKAL), each used by a different laboratory participant. In addition, all participants produced spreadsheet solutions for the test case to check all the mass flows and reactor/facility profiles on a year-by-yearmore » basis throughout the simulation period. The test case specifications describe a transition from the current US fleet of light water reactors to a future fleet of sodium-cooled fast reactors that continuously recycle transuranic elements as fuel. After several initial coordinated modeling and calculation attempts, it was revealed that most of the differences in code results were not due to different code algorithms or calculation approaches, but due to different interpretations of the input specifications among the analysts. Therefore, the specifications for the test case itself were iteratively updated to remove ambiguity and to help calibrate interpretations. In addition, a few corrections and modifications were made to the codes as well, which led to excellent agreement between all codes and spreadsheets for this test case. Although no fuel cycle transition analysis codes matched the spreadsheet results exactly, all remaining differences in the results were due to fundamental differences in code structure and/or were thoroughly explained. As a result, the specifications and example results are provided so that they can be used to verify additional codes in the future for such fuel cycle transition scenarios.« less

Effect of Several Factors on the Cooling of a Radial Engine in Flight

NASA Technical Reports Server (NTRS)

Schey, Oscar W; Pinkel, Benjamin

1936-01-01

Flight tests of a Grumman Scout (XSF-2) airplane fitted with a Pratt & Whitney 1535 supercharged engine were conducted to determine the effect of engine power, mass flow of the cooling air, and atmospheric temperature on cylinder temperature. The tests indicated that the difference in temperature between the cylinder wall and the cooling air varied as the 0.38 power of the brake horsepower for a constant mass flow of cooling air, cooling-air temperature, engine speed, and brake fuel consumption. The difference in temperature was also found to vary inversely as the 0.39 power of the mass flow for points on the head and the 0.35 power for points on the barrel, provided that engine power, engine speed, brake fuel consumption, and cooling-air temperature were kept constant. The results of the tests of the effect of atmospheric temperature on cylinder temperature were inconclusive owing to unfavorable weather conditions prevailing at the time of the tests. The method used for controlling the test conditions, however, was found to be feasible.

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

DOEpatents

Szilard, L.; Young, G.J.

1958-03-01

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

A Burke-Schumann Analysis of Dual-Flame Structure Supported by a Burning Droplet

NASA Technical Reports Server (NTRS)

Nayagam, V.; Dietrich, D.; Williams, F. A.

2016-01-01

Droplet combustion experiments carried out onboard the International Space Station (ISS), using pure fuels and fuel mixtures, have shown that quasi-steady burning can be sustained by a non-traditional flame configuration, namely a "cool flame" burning in the "partial-burning" regime where both fuel and oxygen leak through the low-temperature controlled flame-sheet. Recent experiments involving large, bi-component fuel (n-decane and hexanol, 50/50 by volume) droplets at elevated pressures show that the visible, hot flame becomes extremely weak while the burning rate remains relatively high, suggesting the possibility of simultaneous presence of "cool" and "hot" flames of roughly equal importance. The radiant output from these bi-component droplets is relatively high and cannot be accounted for only by the presence of a visible hot-flame. In this analysis we explore the theoretical possibility of a dual-flame structure, where one flame lies close to the droplet surface called the "cool-flame," and other farther away from the droplet surface, termed the "hot-flame." A Burke-Schumann analysis of this dual-structure seems to indicate such flame structures are possible over a narrow range of initial conditions. Theoretical results can be compared against available experimental data for pure and bi-component fuel droplet combustion to test how realistic the model may be.

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

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

Montierth, Leland M.

2016-07-19

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

Influence of the ambient temperature on the cooling efficiency of the high performance cooling device with thermosiphon effect

NASA Astrophysics Data System (ADS)

Nemec, Patrik; Malcho, Milan

2018-06-01

This work deal with experimental measurement and calculation cooling efficiency of the cooling device working with a heat pipe technology. The referred device in the article is cooling device capable transfer high heat fluxes from electric elements to the surrounding. The work contain description, working principle and construction of cooling device. The main factor affected the dissipation of high heat flux from electronic elements through the cooling device to the surrounding is condenser construction, its capacity and option of heat removal. Experimental part describe the measuring method cooling efficiency of the cooling device depending on ambient temperature in range -20 to 40°C and at heat load of electronic components 750 W. Measured results are compared with results calculation based on physical phenomena of boiling, condensation and natural convection heat transfer.

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

49 CFR 579.21 - Reporting requirements for manufacturers of 5,000 or more light vehicles annually.

Code of Federal Regulations, 2013 CFR

2013-10-01

..., 05 parking brake, 06 engine and engine cooling system, 07 fuel system, 10 power train, 11 electrical... model, the model year, the type, the platform, the fuel and/or propulsion system type coded as follows: CNG (compressed natural gas), CIF (compression ignition fuel), EBP (electric battery power), FCP (fuel...

International Experience in Developing Low-Emission Combustors for Land-Based, Large Gas-Turbine Units: Mitsubishi Heavy Industries' Equipment

NASA Astrophysics Data System (ADS)

Bulysova, L. A.; Vasil'ev, V. D.; Berne, A. L.; Gutnik, M. N.; Ageev, A. V.

2018-05-01

This is the second paper in a series of publications summarizing the international experience in the development of low-emission combustors (LEC) for land-based, large (above 250 MW) gas-turbine units (GTU). The purpose of this series is to generalize and analyze the approaches used by various manufacturers in designing flowpaths for fuel and air in LECs, managing fuel combustion, and controlling the fuel flow. The efficiency of advanced GTUs can be as high as 43% (with an output of 350-500 MW) while the efficiency of 600-800 MW combined-cycle units with these GTUs can attain 63.5%. These high efficiencies require a compression ratio of 20-24 and a temperature as high as 1600°C at the combustor outlet. Accordingly, the temperature in the combustion zone also rises. All the requirements for the control of harmful emissions from these GTUs are met. All the manufacturers and designers of LECs for modern GTUs encounter similar problems, such as emissions control, combustion instability, and reliable cooling of hot path parts. Methods of their elimination are different and interesting from the standpoint of science and practice. One more essential requirement is that the efficiency and environmental performance indices must be maintained irrespective of the fuel composition or heating value and also in operation at part loads below 40% of rated. This paper deals with Mitsubishi Series M701 GTUs, F, G, or J class, which have gained a good reputation in the power equipment market. A design of a burner for LECs and a control method providing stable low-emission fuel combustion are presented. The advantages and disadvantages of the use of air bypass valves installed in each liner to maintain a nearly constant air to fuel ratio within a wide range of GTU loads are described. Methods for controlling low- and high-frequency combustion instabilities are outlined. Upgrading of the cooling system for the wall of a liner and a transition piece is of great interest. Change over from effusion (or film) cooling to convective steam cooling and convective air cooling has considerably increased the GTU efficiency.

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

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

A simulation for predicting potential cooling effect on LPG-fuelled vehicles

NASA Astrophysics Data System (ADS)

Setiyo, M.; Soeparman, S.; Wahyudi, S.; Hamidi, N.

2016-03-01

Liquefied Petroleum Gas vehicles (LPG Vehicles) provide a potential cooling effect about 430 kJ/kg LPG consumption. This cooling effect is obtained from the LPG phase change from liquid to vapor in the vaporizer. In the existing system, energy to evaporate LPG is obtained from the coolant which is circulated around the vaporizer. One advantage is that the LPG (70/30 propane / butane) when expanded from 8 bar to at 1.2 bar, the temperature is less than -25 °C. These conditions provide opportunities to evaporate LPG with ambient air flow, then produce a cooling effect for cooling car's cabin. In this study, some LPG mix was investigated to determine the optimum condition. A simulation was carried out to estimate potential cooling effects of 2000 cc engine from 1000 rpm to 6000 rpm. In this case, the mass flow rate of LPG is a function of fuel consumption. The simulation result shows that the LPG (70/30 propane/butane) provide the greatest cooling effect compared with other mixtures. In conclusion, the 2000 cc engine fueled LPG at 3000 rpm provides potential cooling effect more than 1.3 kW, despite in the low engine speed (1000 rpm) only provides about 0.5 kW.

Can storage reduce electricity consumption? A general equation for the grid-wide efficiency impact of using cooling thermal energy storage for load shifting

NASA Astrophysics Data System (ADS)

Deetjen, Thomas A.; Reimers, Andrew S.; Webber, Michael E.

2018-02-01

This study estimates changes in grid-wide, energy consumption caused by load shifting via cooling thermal energy storage (CTES) in the building sector. It develops a general equation for relating generator fleet fuel consumption to building cooling demand as a function of ambient temperature, relative humidity, transmission and distribution current, and baseline power plant efficiency. The results present a graphical sensitivity analysis that can be used to estimate how shifting load from cooling demand to cooling storage could affect overall, grid-wide, energy consumption. In particular, because power plants, air conditioners and transmission systems all have higher efficiencies at cooler ambient temperatures, it is possible to identify operating conditions such that CTES increases system efficiency rather than decreasing it as is typical for conventional storage approaches. A case study of the Dallas-Fort Worth metro area in Texas, USA shows that using CTES to shift daytime cooling load to nighttime cooling storage can reduce annual, system-wide, primary fuel consumption by 17.6 MWh for each MWh of installed CTES capacity. The study concludes that, under the right circumstances, cooling thermal energy storage can reduce grid-wide energy consumption, challenging the perception of energy storage as a net energy consumer.

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

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

Lee, K.; Feltus, M.A.

1994-12-31

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

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-11-07

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

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-01-01

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

Electrochemical machining process for forming surface roughness elements on a gas turbine shroud

DOEpatents

Lee, Ching-Pang; Johnson, Robert Alan; Wei, Bin; Wang, Hsin-Pang

2002-01-01

The back side recessed cooling surface of a shroud defining in part the hot gas path of a turbine is electrochemically machined to provide surface roughness elements and spaces therebetween to increase the heat transfer coefficient. To accomplish this, an electrode with insulating dielectric portions and non-insulating portions is disposed in opposition to the cooling surface. By passing an electrolyte between the cooling surface and electrode and applying an electrical current between the electrode and a shroud, roughness elements and spaces therebetween are formed in the cooling surface in opposition to the insulating and non-insulating portions of the electrode, hence increasing the surface area and heat transfer coefficient of the shroud.

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

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

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

Intelligent Engine Systems: Thermal Management and Advanced Cooling

NASA Technical Reports Server (NTRS)

Bergholz, Robert

2008-01-01

The objective is to provide turbine-cooling technologies to meet Propulsion 21 goals related to engine fuel burn, emissions, safety, and reliability. Specifically, the GE Aviation (GEA) Advanced Turbine Cooling and Thermal Management program seeks to develop advanced cooling and flow distribution methods for HP turbines, while achieving a substantial reduction in total cooling flow and assuring acceptable turbine component safety and reliability. Enhanced cooling techniques, such as fluidic devices, controlled-vortex cooling, and directed impingement jets, offer the opportunity to incorporate both active and passive schemes. Coolant heat transfer enhancement also can be achieved from advanced designs that incorporate multi-disciplinary optimization of external film and internal cooling passage geometry.

Comparative Performance of Engines Using a Carburetor, Manifold Injection, and Cylinder Injection

NASA Technical Reports Server (NTRS)

Schey, Oscar W; Clark, J Denny

1939-01-01

The comparative performance was determined of engines using three methods of mixing the fuel and the air: the use of a carburetor, manifold injection, and cylinder injection. The tests were made of a single-cylinder engine with a Wright 1820-G air-cooled cylinder. Each method of mixing the fuel and the air was investigated over a range of fuel-air ratios from 0.10 to the limit of stable operation and at engine speeds of 1,500 and 1,900 r.p.m. The comparative performance with a fuel-air ratio of 0.08 was investigated for speeds from 1,300 to 1,900 r.p.m. The results show that the power obtained with each method closely followed the volumetric efficiency; the power was therefore the highest with cylinder injection because this method had less manifold restriction. The values of minimum specific fuel consumption obtained with each method of mixing of fuel and air were the same. For the same engine and cooling conditions, the cylinder temperatures are the same regardless of the method used for mixing the fuel and the air.

Analysis Of Technology Options To Reduce The Fuel Consumption Of Idling Trucks

DOT National Transportation Integrated Search

2000-06-01

Long-haul trucks idling overnight consume more than 838 million gallons (20 million barrels) of fuel annually. Idling also emits pollutants. Truck drivers idle their engines primarily to (1) heat or cool the cab and/or sleeper, (2) keep the fuel warm...

Heat transfer correlations for kerosene fuels and mixtures and physical properties for Jet A fuel

NASA Technical Reports Server (NTRS)

Ackerman, G. H.; Faith, L. E.

1972-01-01

Heat transfer correlations are reported for conventional Jet A fuel for both laminar and turbulent flow in circular tubes. Correlations were developed for cooling in turbine engines, but have broader applications in petroleum and chemical processing, and other industrial applications.

Space Shuttle Upgrades Advanced Hydraulic Power System

NASA Technical Reports Server (NTRS)

2004-01-01

Three Auxiliary Power Units (APU) on the Space Shuttle Orbiter each provide 145 hp shaft power to a hydraulic pump which outputs 3000 psi hydraulic fluid to 41 hydraulic actuators. A hydrazine fuel powered APU utilized throughout the Shuttle program has undergone many improvements, but concerns remain with flight safety, operational cost, critical failure modes, and hydrazine related hazards. The advanced hydraulic power system (AHPS), also known as the electric APU, is being evaluated as an upgrade to replace the hydrazine APU. The AHPS replaces the high-speed turbine and hydrazine fuel supply system with a battery power supply and electric motor/pump that converts 300 volt electrical power to 3000 psi hydraulic power. AHPS upgrade benefits include elimination of toxic hydrazine propellant to improve flight safety, reduction in hazardous ground processing operations, and improved reliability. Development of this upgrade provides many interesting challenges and includes development of four hardware elements that comprise the AHPS system: Battery - The battery provides a high voltage supply of power using lithium ion cells. This is a large battery that must provide 28 kilowatt hours of energy over 99 minutes of operation at 300 volts with a peak power of 130 kilowatts for three seconds. High Voltage Power Distribution and Control (PD&C) - The PD&C distributes electric power from the battery to the EHDU. This 300 volt system includes wiring and components necessary to distribute power and provide fault current protection. Electro-Hydraulic Drive Unit (EHDU) - The EHDU converts electric input power to hydraulic output power. The EHDU must provide over 90 kilowatts of stable, output hydraulic power at 3000 psi with high efficiency and rapid response time. Cooling System - The cooling system provides thermal control of the Orbiter hydraulic fluid and EHDU electronic components. Symposium presentation will provide an overview of the AHPS upgrade, descriptions of the four hardware elements, and a summary of development results to date.

Cryogenic pellet production developments for long-pulse plasma operation

NASA Astrophysics Data System (ADS)

Meitner, S. J.; Baylor, L. R.; Combs, S. K.; Fehling, D. T.; McGill, J. M.; Duckworth, R. C.; McGinnis, W. D.; Rasmussen, D. A.

2014-01-01

Long pulse plasma operation on large magnetic fusion devices require multiple forms of cryogenically formed pellets for plasma fueling, on-demand edge localized mode (ELM) triggering, radiative cooling of the divertor, and impurity transport studies. The solid deuterium fueling and ELM triggering pellets can be formed by extrusions created by helium cooled, twin-screw extruder based injection system that freezes deuterium in the screw section. A solenoid actuated cutter mechanism is activated to cut the pellets from the extrusion, inserting them into the barrel, and then fired by the pneumatic valve pulse of high pressure gas. Fuel pellets are injected at a rate up to 10 Hz, and ELM triggering pellets are injected at rates up to 20 Hz. The radiative cooling and impurity transport study pellets are produced by introducing impurity gas into a helium cooled section of a pipe gun where it deposits in-situ. A pneumatic valve is opened and propellant gas is released downstream where it encounters a passive punch which initially accelerates the pellet before the gas flow around the finishes the pellet acceleration. This paper discusses the various cryogenic pellet production techniques based on the twin-screw extruder, pipe gun, and pellet punch designs.

Cryogenic pellet production developments for long-pulse plasma operation

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

Meitner, S. J.; Baylor, L. R.; Combs, S. K.

Long pulse plasma operation on large magnetic fusion devices require multiple forms of cryogenically formed pellets for plasma fueling, on-demand edge localized mode (ELM) triggering, radiative cooling of the divertor, and impurity transport studies. The solid deuterium fueling and ELM triggering pellets can be formed by extrusions created by helium cooled, twin-screw extruder based injection system that freezes deuterium in the screw section. A solenoid actuated cutter mechanism is activated to cut the pellets from the extrusion, inserting them into the barrel, and then fired by the pneumatic valve pulse of high pressure gas. Fuel pellets are injected at amore » rate up to 10 Hz, and ELM triggering pellets are injected at rates up to 20 Hz. The radiative cooling and impurity transport study pellets are produced by introducing impurity gas into a helium cooled section of a pipe gun where it deposits in-situ. A pneumatic valve is opened and propellant gas is released downstream where it encounters a passive punch which initially accelerates the pellet before the gas flow around the finishes the pellet acceleration. This paper discusses the various cryogenic pellet production techniques based on the twin-screw extruder, pipe gun, and pellet punch designs.« less

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

DOEpatents

Stengel, F.G.

1963-12-24

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

Cold Blooded: Evaluating Brain Temperature by MRI During Surface Cooling of Human Subjects.

PubMed

Curran, Eric J; Wolfson, Daniel L; Watts, Richard; Freeman, Kalev

2017-10-01

Targeted temperature management (TTM) confers neurological and survival benefits for post-cardiac arrest patients with return of spontaneous circulation (ROSC) who remain comatose. Specialized equipment for induction of hypothermia is not available in the prehospital setting, and there are no reliable methods for emergency medical services personnel to initiate TTM. We hypothesized that the application of surface cooling elements to the neck will decrease brain temperature and act as initiators of TTM. Magnetic resonance (MR) spectroscopy was used to evaluate the effect of a carotid surface cooling element on brain temperature in healthy adults. Six individuals completed this study. We measured a temperature drop of 0.69 ± 0.38 °C (95% CI) in the cortex of the brain following the application of the cooling element. Application of a room temperature element also caused a measurable decrease in brain temperature of 0.66 ± 0.41 °C (95% CI) which may be attributable to baroreceptor activation. The application of surface cooling elements to the neck decreased brain temperature and may serve as a method to initiate TTM in the prehospital setting.

THE BIGGEST EXPLOSIONS IN THE UNIVERSE. II

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

Whalen, Daniel J.; Smidt, Joseph; Johnson, Jarrett L.

2013-11-10

One of the leading contenders for the origin of supermassive black holes (SMBHs) at z ∼> 7 is catastrophic baryon collapse in atomically cooled halos at z ∼ 15. In this scenario, a few protogalaxies form in the presence of strong Lyman-Werner UV backgrounds that quench H{sub 2} formation in their constituent halos, preventing them from forming stars or blowing heavy elements into the intergalactic medium prior to formation. At masses of 10{sup 8} M{sub ☉} and virial temperatures of 10{sup 4} K, gas in these halos rapidly cools by H lines, in some cases forming 10{sup 4}-10{sup 6} M{submore » ☉} Population III stars and, a short time later, the seeds of SMBHs. Instead of collapsing directly to black holes (BHs), some of these stars died in the most energetic thermonuclear explosions in the universe. We have modeled the explosions of such stars in the dense cores of line-cooled protogalaxies in the presence of cosmological flows. In stark contrast to the explosions in diffuse regions in previous simulations, these supernovae briefly engulf the protogalaxy, but then collapse back into its dark matter potential. Fallback drives turbulence that efficiently distributes metals throughout the interior of the halo and fuels the rapid growth of nascent BHs at its center. The accompanying starburst and X-ray emission from these line-cooled galaxies easily distinguish them from more slowly evolving neighbors and might reveal the birthplaces of SMBHs on the sky.« less

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

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

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

DOEpatents

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

1961-06-27

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

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-16

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

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

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

SLUG HANDLING DEVICES

DOEpatents

Gentry, J.R.

1958-09-16

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

FUEL ELEMENTS FOR NUCLEAR REACTORS

DOEpatents

Blainey, A.; Lloyd, H.

1961-07-11

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

Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

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

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

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

Evaluation of distributed gas cooling of pressurized PAFC for utility power generation

NASA Technical Reports Server (NTRS)

Farooque, M.; Hooper, M.; Maru, H.

1981-01-01

A proof-of-concept test for a gas-cooled pressurized phosphoric acid fuel cell is described. After initial feasibility studies in short stacks, two 10 kW stacks are tested. Progress includes: (1) completion of design of the test stations with a recirculating gas cooling loop; (2) atmospheric testing of the baseline stack.

Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions

NASA Astrophysics Data System (ADS)

Åberg Lindell, M.; Andersson, P.; Grape, S.; Håkansson, A.; Thulin, M.

2018-07-01

In addition to verifying operator declared parameters of spent nuclear fuel, the ability to experimentally infer such parameters with a minimum of intrusiveness is of great interest and has been long-sought after in the nuclear safeguards community. It can also be anticipated that such ability would be of interest for quality assurance in e.g. recycling facilities in future Generation IV nuclear fuel cycles. One way to obtain information regarding spent nuclear fuel is to measure various gamma-ray intensities using high-resolution gamma-ray spectroscopy. While intensities from a few isotopes obtained from such measurements have traditionally been used pairwise, the approach in this work is to simultaneously analyze correlations between all available isotopes, using multivariate analysis techniques. Based on this approach, a methodology for inferring burnup, cooling time, and initial fissile content of PWR fuels using passive gamma-ray spectroscopy data has been investigated. PWR nuclear fuels, of UOX and MOX type, and their gamma-ray emissions, were simulated using the Monte Carlo code Serpent. Data comprising relative isotope activities was analyzed with decision trees and support vector machines, for predicting fuel parameters and their associated uncertainties. From this work it may be concluded that up to a cooling time of twenty years, the 95% prediction intervals of burnup, cooling time and initial fissile content could be inferred to within approximately 7 MWd/kgHM, 8 months, and 1.4 percentage points, respectively. An attempt aiming to estimate the plutonium content in spent UOX fuel, using the developed multivariate analysis model, is also presented. The results for Pu mass estimation are promising and call for further studies.

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

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

Favalli, Andrea; Vo, D.; Grogan, Brandon R.

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE PAGES

Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

2016-02-26

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Fuel element concept for long life high power nuclear reactors

NASA Technical Reports Server (NTRS)

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

1969-01-01

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

Operating Temperatures of a Sodium-Cooled Exhaust Valve as Measured by a Thermocouple

NASA Technical Reports Server (NTRS)

Sanders, J C; Wilsted, H D; Mulcahy, B A

1943-01-01

Report presents the results of a thermocouple installed in the crown of a sodium-cooled exhaust valve. The valve was tested in an air-cooled engine cylinder and valve temperatures under various engine operating conditions were determined. A temperature of 1337 degrees F. was observed at a fuel-air ratio of 0.064, a brake mean effective pressure of 179 pounds per square inch, and an engine speed of 2000 r.p.m. Fuel-air ratio was found to have a large influence on valve temperature, but cooling-air pressure and variation in spark advance had little effect. An increase in engine power by change of speed or mean effective pressure increased the valve temperature. It was found that the temperature of the rear-spark-plug bushing was not a satisfactory indication of the temperature of the exhaust valve.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

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

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

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

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

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

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

Carlson, P.A.

1963-08-06

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

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Phosphoric Acid Fuel Cell Technology Status

NASA Technical Reports Server (NTRS)

Simons, S. N.; King, R. B.; Prokopius, P. R.

1981-01-01

A review of the current phosphoric acid fuel cell system technology development efforts is presented both for multimegawatt systems for electric utility applications and for multikilowatt systems for on-site integrated energy system applications. Improving fuel cell performance, reducing cost, and increasing durability are the technology drivers at this time. Electrodes, matrices, intercell cooling, bipolar/separator plates, electrolyte management, and fuel selection are discussed.

Design and construction of gas-fed burners for laboratory studies of flame structure

Treesearch

Dan Jimenez; Mark A. Finney; Jack Cohen

2010-01-01

The study of buoyant convection for diffusion flames in wildland fires is critical to understanding heating and cooling dynamics related to particle ignition. Studies based on solid biomass fuels are made difficult by short flame residence time associated with fine fuels. An alternative is to use artificial fuel gas rather than relying on pyrolysis of solid fuels to...

Interrelation of exhaust-gas constituents

NASA Technical Reports Server (NTRS)

Gerrish, Harold C; Voss, Fred

1938-01-01

This report presents the results of an investigation conducted to determine the interrelation of the constituents of the exhaust gases of internal-combustion engines and the effect of engine performance on these relations. Six single-cylinder, liquid-cooled tests engines and one 9-cylinder radial air-cooled engine were tested. Various types of combustion chambers were used and the engines were operated at compression ratios from 5.1 to 7.0 using spark ignition and from 13.5 to 15.6 using compression ignition. The investigation covered a range of engine speeds from 1,500 to 2,100 r.p.m. The fuels used were two grades of aviation gasoline, auto diesel fuel, and laboratory diesel fuel. Power, friction, and fuel-consumption data were obtained from the single-cylinder engines at the same time that the exhaust-gas samples were collected.

Power flattening on modified CANDLE small long life gas-cooled fast reactor

NASA Astrophysics Data System (ADS)

Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

2014-09-01

Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

Fibre optic sensor based measurements of flow-induced vibration in a liquid metal cooled nuclear reactor set-up

NASA Astrophysics Data System (ADS)

De Pauw, B.; Vanlanduit, S.; Van Tichelen, K.; Geernaert, T.; Thienpont, H.; Berghmans, F.

2017-04-01

Fuel assembly vibrations in nuclear reactor cores should not be excessive as these can compromise the lifetime of the assembly and lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. We therefore demonstrate accurate measurements of the vibrations of a fuel assembly in a lead-bismuth eutectic cooled installation with fibre Bragg grating (FBG) based sensors. The use of FBGs in combination with a dedicated sensor integration approach allows accounting for the severe geometrical constraints and providing for the required minimal intrusiveness of the instrumentation, identifying the vibration modes with required accuracy and observing the differences between the vibration amplitudes of the individual fuel pins as well as evidencing a low frequency fuel pin vibration mode resulting from the supports.

Rack for storing spent nuclear fuel elements

DOEpatents

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

1978-06-20

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

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

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

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

DOEpatents

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

1973-12-11

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

Chemical Technology Division annual technical report, 1992

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

Battles, J.E.; Myles, K.M.; Laidler, J.J.

1993-06-01

In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for fluidized-bed combustion and coal-fired magnetohydrodynamics; (3) methods for treatment of hazardous waste, mixed hazardous/radioactive waste, and municipal solid waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for an unsaturated repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, treating water contaminated with volatile organics, and concentrating radioactive waste streams; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (EFR); (7)more » processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials (corium; Fe-U-Zr, tritium in LiAlO{sub 2} in environments simulating those of fission and fusion energy systems. The Division also conducts basic research in catalytic chemistry associated with molecular energy resources and novel` ceramic precursors; materials chemistry of superconducting oxides, electrified metal/solution interfaces, and molecular sieve structures; and the geochemical processes involved in water-rock interactions occurring in active hydrothermal systems. In addition, the Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the technical programs at Argonne National Laboratory (ANL).« less

Interim status report on lead-cooled fast reactor (LFR) research and development.

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

Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

2008-03-31

This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

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

NASA Technical Reports Server (NTRS)

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

2014-01-01

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

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

NASA Technical Reports Server (NTRS)

Ko, William L.

2004-01-01

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

BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

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

T.L. Lotz

1997-02-15

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercialmore » spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.« less

Searching for 300, 000 Degree Gas in the Core of the Phoenix Cluster with HST-COS

NASA Astrophysics Data System (ADS)

McDonald, Michael

2013-10-01

The high central density of the intracluster medium in some galaxy clusters suggests that the hot 10,000,000K gas should cool completely in less than a Hubble time. In these clusters, simple cooling models predict 100-1000 solar masses per year of cooling gas should fuel massive starbursts in the central galaxy. The fact that the typical central cluster galaxy is a massive, "red and dead" elliptical galaxy, with little evidence for a cool ISM, has led to the realization of the "cooling flow problem". It is now thought that mechanical feedback from the central supermassive blackhole, in the form of radio-blown bubbles, is offsetting cooling, leading to an exceptionally precise {residuals of less than 10 percent} balance between cooling and feedback in nearly every galaxy cluster in the local Universe. In the recently-discovered Phoenix cluster, where z=0.596, we observe an 800 solar mass per year starburst within the central galaxy which accounts for about 30 percent of the classical cooling prediction for this system. We speculate that this may represent the first "true" cooling flow, with the factor of 3 difference between cooling and star formation being attributed to star formation efficiency, rather than a problem with cooling. In order to test these predictions, we propose far-UV spectroscopic observations of the OVI 1032A emission line, which probes 10^5.5K gas, in the central galaxy of the Phoenix cluster. If detected at the expected levels, this would provide compelling evidence that the starburst is, indeed, fueled by runaway cooling of the intracluster medium, confirming the presence of the first, bonafide cooling flow.

NASA Reactor Facility Hazards Summary. Volume 1

NASA Technical Reports Server (NTRS)

1959-01-01

The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

NASA Technical Reports Server (NTRS)

Walton, J. T.

1994-01-01

ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

Fuel cell crimp-resistant cooling device with internal coil

NASA Technical Reports Server (NTRS)

Wittel, deceased, Charles F. (Inventor)

1986-01-01

A cooling assembly for fuel cells having a simplified construction whereby coolant is efficiently circulated through a conduit arranged in serpentine fashion in a channel within a member of such assembly. The channel is adapted to cradle a flexible, chemically inert, conformable conduit capable of manipulation into a variety of cooling patterns without crimping or otherwise restricting of coolant flow. The conduit, when assembled with the member, conforms into intimate contact with the member for good thermal conductivity. The conduit is non-corrodible and can be constructed as a single, manifold-free, continuous coolant passage means having only one inlet and one outlet. The conduit has an internal coil means which enables it to be bent in small radii without crimping.

A fuselage/tank structure study for actively cooled hypersonic cruise vehicles: Active cooling system analysis

NASA Technical Reports Server (NTRS)

Stone, J. E.

1975-01-01

The effects of fuselage cross section and structural arrangement on the performance of actively cooled hypersonic cruise vehicles are investigated. An active cooling system which maintains the aircraft's entire surface area at temperatures below 394 K at Mach 6 is developed along with a hydrogen fuel tankage thermal protection system. Thermodynamic characteristics of the actively cooled thermal protection systems established are summarized. Design heat loads and coolant flowrate requirements are defined for each major structural section and for the total system. Cooling system weights are summarized at the major component level. Conclusions and recommendations are included.

Nuclear fuel element

DOEpatents

Zocher, Roy W.

1991-01-01

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

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

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

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

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

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

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

1991-01-01

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

TECHNICAL SCOPE OF GAS-COOLED REACTOR FUEL ELEMENT IRRADIATION PROGRAM

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

None

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

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

Dynamic model of a micro-tubular solid oxide fuel cell stack including an integrated cooling system

NASA Astrophysics Data System (ADS)

Hering, Martin; Brouwer, Jacob; Winkler, Wolfgang

2017-02-01

A novel dynamic micro-tubular solid oxide fuel cell (MT-SOFC) and stack model including an integrated cooling system is developed using a quasi three-dimensional, spatially resolved, transient thermodynamic, physical and electrochemical model that accounts for the complex geometrical relations between the cells and cooling-tubes. The modeling approach includes a simplified tubular geometry and stack design including an integrated cooling structure, detailed pressure drop and gas property calculations, the electrical and physical constraints of the stack design that determine the current, as well as control strategies for the temperature. Moreover, an advanced heat transfer balance with detailed radiative heat transfer between the cells and the integrated cooling-tubes, convective heat transfer between the gas flows and the surrounding structures and conductive heat transfer between the solid structures inside of the stack, is included. The detailed model can be used as a design basis for the novel MT-SOFC stack assembly including an integrated cooling system, as well as for the development of a dynamic system control strategy. The evaluated best-case design achieves very high electrical efficiency between around 75 and 55% in the entire power density range between 50 and 550 mW /cm2 due to the novel stack design comprising an integrated cooling structure.

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

Konyashov, Vadim V.; Krasnov, Alexander M.

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

Dry-vault storage of spent fuel at the CASCAD facility

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

Baillif, L.; Guay, M.

A new modular dry storage vault concept using vertical metallic wells cooled by natural convection has been developed by the Commissariat a l'Energie Atomique and Societe Generale pour les Techniques Nouvelles to accommodate special fuels for high-level wastes. Basic specifications and design criteria have been followed to guarantee a double containment system and cooling to maintain the fuel below an acceptable temperature. The double containment is provided by two static barriers: At the reactor, fuels are placed in containers playing the role of the first barrier; the storage wells constitute the second barrier. Spent fuel placed in wells is cooledmore » by natural convection: a boundary layer is created along the outer side of the well. The heated air rises along the well leading to a thermosiphon flow that extracts the heat released. For heat transfer, studies, computations, and experimental tests have been carried out to calculate and determine the temperature of the containers and the fuel rod temperatures in various situations. The CASCAD vault storage can be applied to light water reactor (LWR) fuels without any difficulties if two requirements are satisfied: (1) Spend fuels have to be inserted in tight canisters. (2) Spent fuels have to be received only after a minimum decay time of 5 yr.« less

Refuse-Derived Fuels in U.S. Air Force Heating and Power Systems.

DTIC Science & Technology

1986-01-01

pellet degradation . Postpellet - cooling may not prevent disintegration into more fines. The Argonne W National Laboratory is sponsoring a research...of fuel,% by using the fuel Orsat (volumetric) analysis and both the carbon and sulfur in the fuel. This is known as a carbon balance. 00 The test... sulfur in the fuel per pound of fuel C ’ = average specific heat of the dry flue gas PG T = flue gas temperature leaving the economizer (or leaving

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)

14 CFR 33.67 - Fuel system.

Code of Federal Regulations, 2012 CFR

2012-01-01

... STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.67 Fuel system. (a) With... range with the fuel initially saturated with water at 80 °F (27 °C) and having 0.025 fluid ounces per gallon (0.20 milliliters per liter) of free water added and cooled to the most critical condition for...

Photographic combustion characterization of LOX/Hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1980-01-01

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

Development, Testing and Validation of a Waste Assay System for the Measurement and Characterisation of Active Spent Fuel Element Debris From UK Magnox Reactors - 12533

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

Mason, John A.; Burke, Kevin J.; Looman, Marc R.

2012-07-01

This paper describes the development, testing and validation of a waste measurement instrument for characterising active remote handled radioactive waste arising from the operation of Magnox reactors in the United Kingdom. Following operation in UK Magnox gas cooled reactors and a subsequent period of cooling, parts of the magnesium-aluminium alloy cladding were removed from spent fuel and the uranium fuel rods with the remaining cladding were removed to Sellafield for treatment. The resultant Magnox based spent fuel element debris (FED), which constitutes active intermediate level waste (ILW) has been stored in concrete vaults at the reactor sites. As part ofmore » the decommissioning of the FED vaults the FED must be removed, measured and characterised and placed in intermediate storage containers. The present system was developed for use at the Trawsfynydd nuclear power station (NPS), which is in the decommissioning phase, but the approach is potentially applicable to FED characterisation at all of the Magnox reactors. The measurement system consists of a heavily shielded and collimated high purity Germanium (HPGe) detector with electromechanical cooling and a high count-rate preamplifier and digital multichannel pulse height analyser. The HPGe based detector system is controlled by a software code, which stores the measurement result and allows a comprehensive analysis of the measured FED data. Fuel element debris is removed from the vault and placed on a tray to a uniform depth of typically 10 cm for measurement. The tray is positioned approximately 1.2 meters above the detector which views the FED through a tungsten collimator with an inverted pyramid shape. At other Magnox sites the positions may be reversed with the shielded and collimated HPGe detector located above the tray on which the FED is measured. A comprehensive Monte Carlo modelling and analysis of the measurement process has been performed in order to optimise the measurement geometry and eliminate interferences from radioactive sources and FED in the immediate vicinity of the measurement position. The detector system has been calibrated and high activity radioactive sources of Cs-137, Co-60 and Na-22 have been used to validate the measurement process. The data acquisition and analysis software code has been tested and validated in keeping with the software quality assurance requirements of both ISO:9001-2008 - TICK-IT in the UK and NQA-1. The measurement and analysis system has been comprehensively tested with high activity sources, is flexible and may be applicable to a wide range of remote handled radioactive waste measurement applications. It is due to be installed at Trawsfynydd NPS later this year. This paper describes the Waste Tray Assay System (WTAS) that has been developed for the measurement of Magnox FED waste. The WTAS has been tested with a range of radioactive sources and its operation has been simulated with benchmarked MCNP Monte Carlo calculations. The measurement software has been validated as has the operation of the system for a range of strong radioactive sources. A system based on the design is due for installation and operation in 2012. The system has application to the measurement of Magnox Fuel Element Debris (FED) waste at other Magnox reactor sites. The major design objective of the WTAS that has been achieved is the ability of the assay system to determine the content of Cs-137, and in turn to enable the fissile burden to be assessed using a radionuclide fingerprint, in the presence of higher and highly variable quantities of Co-60, typically from nimonic springs. The approach can be used in other Magnox FED waste configurations where the detector is located above the FED waste sorting tray and where the collimation is fixed below the detector and at a distance above the tray. In this case, which has also been investigated, there are different shielding problems and mechanical support issues. The extensive use of MCNP Monte Carlo modelling to simulate the geometry of the sorting cell and the distribution of radioactive sources has helped to ensure that all of the detector shielding requirements are addressed and suitable Cs-137 and Co-60 discrimination can be achieved. The WTAS in its present form or in other configurations has relevance to the measurement of other active ILW and highly active RH waste. Examples include high activity RH LLW and RH TRU (Transuranic) waste as defined in the United States arising from both commercial nuclear and Department of Energy (DOE) operations. The analysis is able to analyse a range of radionuclides beyong those expected in the Magnox FED cases. (authors)« less

Pumpless thermal management of water-cooled high-temperature proton exchange membrane fuel cells

NASA Astrophysics Data System (ADS)

Song, Tae-Won; Choi, Kyoung-Hwan; Kim, Ji-Rae; Yi, Jung S.

2011-05-01

Proton exchange membrane fuel cells (PEMFCs) have been considered for combined heat and power (CHP) applications, but cost reduction has remained an issue for commercialization. Among various types of PEMFC, the high-temperature (HT) PEMFC is gaining more attention due to the simplicity of the system, that will make the total system cost lower. A pumpless cooling concept is introduced to reduce the number of components of a HT PEMFC system even further and also decrease the parasitic power required for operating the system. In this concept, water is used as the coolant, and the buoyancy force caused by the density difference between vapour and liquid when operated above boiling temperate is utilized to circulate the coolant between the stack and the cooling device. In this study, the basic parameters required to design the cooling device are discussed, and the stable operation of the HT PEMFC stack in both the steady-state and during transient periods is demonstrated. It found that the pumpless cooling method provides more uniform temperature distribution within the stack, regardless of the direction of coolant flow.

Fuel handling apparatus for a nuclear reactor

DOEpatents

Hawke, Basil C.

1987-01-01

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

Zirconium determination by cooling curve analysis during the pyroprocessing of used nuclear fuel

NASA Astrophysics Data System (ADS)

Westphal, B. R.; Price, J. C.; Bateman, K. J.; Marsden, K. C.

2015-02-01

An alternative method to sampling and chemical analyses has been developed to monitor the concentration of zirconium in real-time during the casting of uranium products from the pyroprocessing of used nuclear fuel. The method utilizes the solidification characteristics of the uranium products to determine zirconium levels based on standard cooling curve analyses and established binary phase diagram data. Numerous uranium products have been analyzed for their zirconium content and compared against measured zirconium data. From this data, the following equation was derived for the zirconium content of uranium products:

Removal of hydrogen sulfide as ammonium sulfate from hydropyrolysis product vapors

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

Marker, Terry L.; Felix, Larry G.; Linck, Martin B.

A system and method for processing biomass into hydrocarbon fuels that includes processing a biomass in a hydropyrolysis reactor resulting in hydrocarbon fuels and a process vapor stream and cooling the process vapor stream to a condensation temperature resulting in an aqueous stream. The aqueous stream is sent to a catalytic reactor where it is oxidized to obtain a product stream containing ammonia and ammonium sulfate. A resulting cooled product vapor stream includes non-condensable process vapors comprising H.sub.2, CH.sub.4, CO, CO.sub.2, ammonia and hydrogen sulfide.

Removal of hydrogen sulfide as ammonium sulfate from hydropyrolysis product vapors

DOEpatents

Marker, Terry L; Felix, Larry G; Linck, Martin B; Roberts, Michael J

2014-10-14

A system and method for processing biomass into hydrocarbon fuels that includes processing a biomass in a hydropyrolysis reactor resulting in hydrocarbon fuels and a process vapor stream and cooling the process vapor stream to a condensation temperature resulting in an aqueous stream. The aqueous stream is sent to a catalytic reactor where it is oxidized to obtain a product stream containing ammonia and ammonium sulfate. A resulting cooled product vapor stream includes non-condensable process vapors comprising H.sub.2, CH.sub.4, CO, CO.sub.2, ammonia and hydrogen sulfide.

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

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

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

1998-06-01

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

Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

Thermostructural applications of heat pipes for cooling leading edges of high-speed aerospace vehicles

NASA Technical Reports Server (NTRS)

Camarda, Charles J.; Glass, David E.

1992-01-01

Heat pipes have been considered for use on wing leading edge for over 20 years. Early concepts envisioned metal heat pipes cooling a metallic leading edge. Several superalloy/sodium heat pipes were fabricated and successfully tested for wing leading edge cooling. Results of radiant heat and aerothermal testing indicate the feasibility of using heat pipes to cool the stagnation region of shuttle-type space transportation systems. The test model withstood a total seven radiant heating tests, eight aerothermal tests, and twenty-seven supplemental radiant heating tests. Cold-wall heating rates ranged from 21 to 57 Btu/sq ft-s and maximum operating temperatures ranged from 1090 to 1520 F. Follow-on studies investigated the application of heat pipes to cool the stagnation regions of single-stage-to-orbit and advanced shuttle vehicles. Results of those studies indicate that a 'D-shaped' structural design can reduce the mass of the heat-pipe concept by over 44 percent compared to a circular heat-pipe geometry. Simple analytical models for heat-pipe startup from the frozen state (working fluid initially frozen) were adequate to approximate transient, startup, and steady-state heat-pipe performance. Improvement in analysis methods has resulted in the development of a finite-element analysis technique to predict heat-pipe startup from the frozen state. However, current requirements of light-weight design and reliability suggest that metallic heat pipes embedded in a refractory composite material should be used. This concept is the concept presently being evaluated for NASP. A refractory-composite/heat-pipe-cooled wing leading edge is currently being considered for the National Aero-Space Plane (NASP). This concept uses high-temperature refractory-metal/lithium heat pipes embedded within a refractory-composite structure and is significantly lighter than an actively cooled wing leading edge because it eliminates the need for active cooling during ascent and descent. Since the NASP vehicle uses cryogenic hydrogen to cool structural components and then burns this fuel in the combustor, hydrogen necessary for descent cooling only, when the vehicle is unpowered, is considered to be a weight penalty. Details of the design of the refractory-composite/heat-pipe-cooled wing leading edge are currently being investigated. Issues such as thermal contact resistance and thermal stress are also being investigated.

Thermostructural applications of heat pipes for cooling leading edges of high-speed aerospace vehicles

NASA Astrophysics Data System (ADS)

Camarda, Charles J.; Glass, David E.

1992-10-01

Heat pipes have been considered for use on wing leading edge for over 20 years. Early concepts envisioned metal heat pipes cooling a metallic leading edge. Several superalloy/sodium heat pipes were fabricated and successfully tested for wing leading edge cooling. Results of radiant heat and aerothermal testing indicate the feasibility of using heat pipes to cool the stagnation region of shuttle-type space transportation systems. The test model withstood a total seven radiant heating tests, eight aerothermal tests, and twenty-seven supplemental radiant heating tests. Cold-wall heating rates ranged from 21 to 57 Btu/sq ft-s and maximum operating temperatures ranged from 1090 to 1520 F. Follow-on studies investigated the application of heat pipes to cool the stagnation regions of single-stage-to-orbit and advanced shuttle vehicles. Results of those studies indicate that a 'D-shaped' structural design can reduce the mass of the heat-pipe concept by over 44 percent compared to a circular heat-pipe geometry. Simple analytical models for heat-pipe startup from the frozen state (working fluid initially frozen) were adequate to approximate transient, startup, and steady-state heat-pipe performance. Improvement in analysis methods has resulted in the development of a finite-element analysis technique to predict heat-pipe startup from the frozen state. However, current requirements of light-weight design and reliability suggest that metallic heat pipes embedded in a refractory composite material should be used. This concept is the concept presently being evaluated for NASP. A refractory-composite/heat-pipe-cooled wing leading edge is currently being considered for the National Aero-Space Plane (NASP). This concept uses high-temperature refractory-metal/lithium heat pipes embedded within a refractory-composite structure and is significantly lighter than an actively cooled wing leading edge because it eliminates the need for active cooling during ascent and descent. Since the NASP vehicle uses cryogenic hydrogen to cool structural components and then burns this fuel in the combustor, hydrogen necessary for descent cooling only, when the vehicle is unpowered, is considered to be a weight penalty. Details of the design of the refractory-composite/heat-pipe-cooled wing leading edge are currently being investigated. Issues such as thermal contact resistance and thermal stress are also being investigated.

The supersonic molecular beam injector as a reliable tool for plasma fueling and physics experiment on HL-2A.

PubMed

Chen, C Y; Yu, D L; Feng, B B; Yao, L H; Song, X M; Zang, L G; Gao, X Y; Yang, Q W; Duan, X R

2016-09-01

On HL-2A tokamak, supersonic molecular beam injection (SMBI) has been developed as a routine refueling method. The key components of the system are an electromagnetic valve and a conic nozzle. The valve and conic nozzle are assembled to compose the simplified Laval nozzle for generating the pulsed beam. The appurtenance of the system includes the cooling system serving the cooled SMBI generation and the in situ calibration component for quantitative injection. Compared with the conventional gas puffing, the SMBI features prompt response and larger fueling flux. These merits devote the SMBI a good fueling method, an excellent plasma density feedback control tool, and an edge localized mode mitigation resource.

Advanced high pressure engine study for mixed-mode vehicle applications

NASA Technical Reports Server (NTRS)

Luscher, W. P.; Mellish, J. A.

1977-01-01

High pressure liquid rocket engine design, performance, weight, envelope, and operational characteristics were evaluated for a variety of candidate engines for use in mixed-mode, single-stage-to-orbit applications. Propellant property and performance data were obtained for candidate Mode 1 fuels which included: RP-1, RJ-5, hydrazine, monomethyl-hydrazine, and methane. The common oxidizer was liquid oxygen. Oxygen, the candidate Mode 1 fuels, and hydrogen were evaluated as thrust chamber coolants. Oxygen, methane, and hydrogen were found to be the most viable cooling candidates. Water, lithium, and sodium-potassium were also evaluated as auxiliary coolant systems. Water proved to be the best of these, but the system was heavier than those systems which cooled with the engine propellants. Engine weight and envelope parametric data were established for candidate Mode 1, Mode 2, and dual-fuel engines. Delivered engine performance data were also calculated for all candidate Mode 1 and dual-fuel engines.

Beam heated linear theta-pinch device for producing hot plasmas

DOEpatents

Bohachevsky, Ihor O.

1981-01-01

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

A Combined Water-Bromotrifluoromethane Crash-Fire Protection System for a T-56 Turbopropeller Engine

NASA Technical Reports Server (NTRS)

Campbell, John A.; Busch, Arthur M.

1959-01-01

A crash-fire protection system is described which will suppress the ignition of crash-spilled fuel that may be ingested by a T-56 turbo-propeller engine. This system includes means for rapidly extinguishing the combustor flame, means for cooling and inerting with water the hot engine parts likely to ignite engine ingested fuel, and means for blanketing with bromotrifluoromethane massive metal parts that may reheat after the engine stops rotating. Combustion-chamber flames were rapidly extinguished at the engine fuel nozzles by a fuel shutoff and drain valve. Hot engine parts were inerted and cooled by 42 pounds of water discharged at seven engine stations. Massive metal parts that could reheat were inerted with 10 pounds of bromotrifluoromethane discharged at two engine stations. Performance trials of the crash-fire protection system were conducted by bringing the engine up to takeoff temperature, actuating the crash-fire protection system, and then spraying fuel into the engine to simulate crash-ingested fuel. No fires occurred during these trials, although fuel was sprayed into the engine from 0.3 second to 15 minutes after actuating the crash-fire protection system.

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

NASA Astrophysics Data System (ADS)

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

2016-01-01

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

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

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

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

2016-01-22

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

Wright R–2600–8 Engine in the Engine Propeller Research Building

NASA Image and Video Library

1943-03-21

A Wright Aeronautical R–2600 Cyclone piston engine installed in the Engine Propeller Research Building, or Prop House, at the National Advisory Committee for Aeronautics (NACA) Aircraft Engine Research Laboratory. The R–2600 was among the most powerful engines that emerged during World War II. The engine, which was developed for commercial applications in 1939, was used to power the North American B–25 bomber and several other midsize military aircraft. The higher altitudes required by the military caused problems with the engine's cooling and fuel systems. The military requested that the Aircraft Engine Research Laboratory analyze the performance of the R–2600, improve its cooling system, and reduce engine knock. The NACA researchers subjected the engine to numerous tests in its Prop House. The R–2600 was the subject of the laboratory's first technical report, which was written by members of the Fuels and Lubricants Division. The Prop House contained soundproof test cells in which piston engines and propellers were mounted and operated at high powers. Electrically driven fans drew air through ducts to create a stream of cooling air over the engines. Researchers tested the performance of fuels, turbochargers, water-injection and cooling systems here during World War II. The facility was also investigated a captured German V–I buzz bomb during the war.

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

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

Parallel array of independent thermostats for column separations

DOEpatents

Foret, Frantisek; Karger, Barry L.

2005-08-16

A thermostat array including an array of two or more capillary columns (10) or two or more channels in a microfabricated device is disclosed. A heat conductive material (12) surrounded each individual column or channel in array, each individual column or channel being thermally insulated from every other individual column or channel. One or more independently controlled heating or cooling elements (14) is positioned adjacent to individual columns or channels within the heat conductive material, each heating or cooling element being connected to a source of heating or cooling, and one or more independently controlled temperature sensing elements (16) is positioned adjacent to the individual columns or channels within the heat conductive material. Each temperature sensing element is connected to a temperature controller.

Viscoelastic finite element analysis of residual stresses in porcelain-veneered zirconia dental crowns.

PubMed

Kim, Jeongho; Dhital, Sukirti; Zhivago, Paul; Kaizer, Marina R; Zhang, Yu

2018-06-01

The main problem of porcelain-veneered zirconia (PVZ) dental restorations is chipping and delamination of veneering porcelain owing to the development of deleterious residual stresses during the cooling phase of veneer firing. The aim of this study is to elucidate the effects of cooling rate, thermal contraction coefficient and elastic modulus on residual stresses developed in PVZ dental crowns using viscoelastic finite element methods (VFEM). A three-dimensional VFEM model has been developed to predict residual stresses in PVZ structures using ABAQUS finite element software and user subroutines. First, the newly established model was validated with experimentally measured residual stress profiles using Vickers indentation on flat PVZ specimens. An excellent agreement between the model prediction and experimental data was found. Then, the model was used to predict residual stresses in more complex anatomically-correct crown systems. Two PVZ crown systems with different thermal contraction coefficients and porcelain moduli were studied: VM9/Y-TZP and LAVA/Y-TZP. A sequential dual-step finite element analysis was performed: heat transfer analysis and viscoelastic stress analysis. Controlled and bench convection cooling rates were simulated by applying different convective heat transfer coefficients 1.7E-5 W/mm 2 °C (controlled cooling) and 0.6E-4 W/mm 2 °C (bench cooling) on the crown surfaces exposed to the air. Rigorous viscoelastic finite element analysis revealed that controlled cooling results in lower maximum stresses in both veneer and core layers for the two PVZ systems relative to bench cooling. Better compatibility of thermal contraction coefficients between porcelain and zirconia and a lower porcelain modulus reduce residual stresses in both layers. Copyright © 2018 Elsevier Ltd. All rights reserved.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

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

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

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

Charpenel, Jean; Lemoine, Francette; Sato, Ikken

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

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

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

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

2015-03-01

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

Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4

NASA Astrophysics Data System (ADS)

Hallman, Luther, Jr.

Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism. One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel. To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior. This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.

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

DOEpatents

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

1982-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

Energy Sustainability and the Army: The Current Transformation

DTIC Science & Technology

2009-04-01

heating and 20 cooling for buildings; wood-fired central heating plant; pyrolysis plant for conversion of wood to liquid fuels; synthetic mobility fuels...Http://Www.Sciencedaily.Com/Releases/2008/ 08/080818184434.Htm. Accessed 22 August 2008. Anonymous, “Bio-Fuel from Corn , Switch-Grass And Misconthus...Press_Releases_Seven.Html. Accessed 15 August 2008. Anonymous, “Protecting Soils and Producing Bio-Fuel With Corn Stover, Science News, 7 November 2008. Available

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

Assemblies with both target and fuel pins in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

Fire resistant nuclear fuel cask

DOEpatents

Heckman, Richard C.; Moss, Marvin

1979-01-01

The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked.

Copper, zinc, gallium and germanium distributions in taenite lamellae of iron meteorites and their importance for cooling rate estimations

NASA Astrophysics Data System (ADS)

Braun-Dullaeus, Karl-Ulrich; Traxel, Kurt

1995-02-01

One method forestimating cooling rates of meteorite parent bodies is to model measured nickel distributions in taenite lamellae of iron meteorites. Goldstein and Ogilvie ( Geochim. Cosmochim. Acta29, 893, 1965) and Rasmussen ( Icarus45, 564, 1981) developed techniques based on this idea to examine the cooling history in the temperature range between ˜700 and ˜400°C. As a result of Instrumental Neutron Activation Analysis (INAA) Rasmussen et al. ( Meteoritics23, 105, 1988) postulated that some trace elements would also be good cooling rate indicators. They argued that elements with distinct diffusion behavior are sensitive to different temperature ranges. The new Heidelberg proton microprobe uses the method of Proton Induced X-ray Emission (PIXE) for elemental analysis. This microprobe is an appropriate instrument to measure distributions of trace elements with a spatial resolution of 2 μm. We demonstrated on the iron meteorites Cape York (Agpalilik), Toluca and Odessa that the elements copper, zinc, gallium and germanium imitate the profiles of nickel in taenite lamella. The interpretation of the Zn, Ga and Ge profiles leads to the conclusion that these elements undergo diffusion mechanisms comparable to those of Ni. The numerical simulation of Cu distributions with a simplified model points out that little new information can be obtained about the cooling history of the meteorites by modelling Cu profiles. To simulate Zn, Ga or Ge distributions, the use of ternary phase diagrams is necessary.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-23

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

Experimental Investigation of Turbine Vane Heat Transfer for Alternative Fuels

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

Nix, Andrew Carl

The focus of this program was to experimentally investigate advanced gas turbine cooling schemes and the effects of and factors that contribute to surface deposition from particulate matter found in coal syngas exhaust flows on turbine airfoil heat transfer and film cooling, as well as to characterize surface roughness and determine the effects of surface deposition on turbine components. The program was a comprehensive, multi-disciplinary collaborative effort between aero-thermal and materials faculty researchers and the Department of Energy, National Energy Technology Laboratory (NETL). The primary technical objectives of the program were to evaluate the effects of combustion of syngas fuelsmore » on heat transfer to turbine vanes and blades in land-based power generation gas turbine engines. The primary questions to be answered by this investigation were; What are the factors that contribute to particulate deposition on film cooled gas turbine components? An experimental program was performed in a high-temperature and pressure combustion rig at the DOE NETL; What is the effect of coal syngas combustion and surface deposition on turbine airfoil film cooling? Deposition of particulate matter from the combustion gases can block film cooling holes, decreasing the flow of the film coolant and the film cooling effectiveness; How does surface deposition from coal syngas combustion affect turbine surface roughness? Increased surface roughness can increase aerodynamic losses and result in decreased turbine hot section efficiency, increasing engine fuel consumption to maintain desired power output. Convective heat transfer is also greatly affected by the surface roughness of the airfoil surface; Is there any significant effect of surface deposition or erosion on integrity of turbine airfoil thermal barrier coatings (TBC) and do surface deposits react with the TBC in any way to decrease its thermal insulating capability? Spallation and erosion of TBC is a persistent problem in modern turbine engines; and What advancements in film cooling hole geometry and design can increase effectiveness of film cooling in turbines burning high-hydrogen coal syngas due to the higher heat loads and mass flow rates of the core flow? Experimental and numerical investigations of advanced cooling geometries that can improve resistance to surface deposition were performed. The answers to these questions were investigated through experimental measurements of turbine blade surface temperature and coolant coverage (via infrared camera images and thermocouples) and time-varying surface roughness in the NETL high-pressure combustion rig with accelerated, simulated surface deposition and advanced cooling hole concepts, coupled with detailed materials analysis and characterization using conventional methods of Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM), X-Ray Diffraction (XRD), 3-D Surface Topography (using a 3-D stylus profilometer). Detailed surface temperatures and cooling effectiveness could not be measured due to issues with the NETL infrared camera system. In collaboration with faculty startup funding from the principal investigator, experimental and numerical investigations were performed of an advanced film cooling hole geometry, the anti-vortex hole (AVH), focusing on improving cooling effectiveness and decreasing the counter-rotating vortex of conventional cooling holes which can entrain mainstream particulate matter to the surface. The potential benefit of this program is in gaining a fundamental understanding of how the use of alternative fuels will effect the operation of modern gas turbine engines, providing valuable data for more effective cooling designs for future turbine systems utilizing alternative fuels.« less

NASA Engineer Examines the Design of a Regeneratively-Cooled Rocket Engine

NASA Image and Video Library

1958-12-21

An engineer at the National Aeronautics and Space Administration (NASA) Lewis Research Center examines a drawing showing the assembly and details of a 20,000-pound thrust regeneratively cooled rocket engine. The engine was being designed for testing in Lewis’ new Rocket Engine Test Facility, which began operating in the fall of 1957. The facility was the largest high-energy test facility in the country that was capable of handling liquid hydrogen and other liquid chemical fuels. The facility’s use of subscale engines up to 20,000 pounds of thrust permitted a cost-effective method of testing engines under various conditions. The Rocket Engine Test Facility was critical to the development of the technology that led to the use of hydrogen as a rocket fuel and the development of lightweight, regeneratively-cooled, hydrogen-fueled rocket engines. Regeneratively-cooled engines use the cryogenic liquid hydrogen as both the propellant and the coolant to prevent the engine from burning up. The fuel was fed through rows of narrow tubes that surrounded the combustion chamber and nozzle before being ignited inside the combustion chamber. The tubes are visible in the liner sitting on the desk. At the time, Pratt and Whitney was designing a 20,000-pound thrust liquid-hydrogen rocket engine, the RL-10. Two RL-10s would be used to power the Centaur second-stage rocket in the 1960s. The successful development of the Centaur rocket and the upper stages of the Saturn V were largely credited to the work carried out Lewis.

THE MANUFACTURE OF FUEL ELEMENTS OF THE ARGONAUT TYPE

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

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

1958-06-10

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

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

NASA Technical Reports Server (NTRS)

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

2013-01-01

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

Technical and economic assessment of different options for minor actinide transmutation: the French case

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

Chabert, C.; Coquelet-Pascal, C.; Saturnin, A.

Studies have been performed to assess the industrial perspectives of partitioning and transmutation of long-lived elements. These studies were carried out in tight connection with GEN-IV systems development. The results include the technical and economic evaluation of fuel cycle scenarios along with different options for optimizing the processes between the minor actinide transmutation in fast neutron reactors, their interim storage and geological disposal of ultimate waste. The results are analysed through several criteria (impacts on waste, on waste repository, on fuel cycle plants, on radiological exposure of workers, on costs and on industrial risks). These scenario evaluations take place inmore » the French context which considers the deployment of the first Sodium-cooled Fast Reactor (SFR) in 2040. 3 management options of minor actinides have been studied: no transmutation, transmutation in SFR and transmutation in an accelerator-driven system (ADS). Concerning economics the study shows that the cost overrun related to the transmutation process could vary between 5 to 9% in SFR and 26 % in the case of ADS.« less

STEAM FORMING NEUTRONIC REACTOR AND METHOD OF OPERATING IT

DOEpatents

Untermyer, S.

1960-05-10

The heterogeneous reactor is liquid moderated and cooled by a steam forming coolant and is designed to produce steam from the coolant directly within the active portion of the reactor while avoiding the formation of bubbles in the liquid moderator. This reactor achieves inherent stability as a result of increased neutron leakage and increased neutron resonance absorption in the U/sup 238/ fuel with the formation of bubbles. The invention produces certain conditions under which the formation of vapor bubbles as a result of a neutron flux excursion from the injection of a reactivity increment into the reactor will operate to nullify the reactivity increment within a sufficiently short period of time to prevent unsafe reactor operating conditions from developing. This is obtained by disposing a plurality of fuel elements within a mass of steam forming coolant in the core with the ratio of the volume of steam forming coolant to the volume of fissionable isotopes being within the range yielding a multiplication factor greater than unity and a negative reactivity to core void coefficient at the boiling temperature of the coolant.

14 CFR 25.1043 - Cooling tests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... (a)(1) of this section may exceed established limits. (3) For reciprocating engines, the fuel used during the cooling tests must be the minimum grade approved for the engines, and the mixture settings... engine fluids and powerplant components (except cylinder barrels) for which temperature limits are...

14 CFR 25.1043 - Cooling tests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... (a)(1) of this section may exceed established limits. (3) For reciprocating engines, the fuel used during the cooling tests must be the minimum grade approved for the engines, and the mixture settings... engine fluids and powerplant components (except cylinder barrels) for which temperature limits are...

Improving Engine Efficiency Through Core Developments

NASA Technical Reports Server (NTRS)

Heidmann, James D.

2011-01-01

The NASA Environmentally Responsible Aviation (ERA) Project and Fundamental Aeronautics Projects are supporting compressor and turbine research with the goal of reducing aircraft engine fuel burn and greenhouse gas emissions. The primary goals of this work are to increase aircraft propulsion system fuel efficiency for a given mission by increasing the overall pressure ratio (OPR) of the engine while maintaining or improving aerodynamic efficiency of these components. An additional area of work involves reducing the amount of cooling air required to cool the turbine blades while increasing the turbine inlet temperature. This is complicated by the fact that the cooling air is becoming hotter due to the increases in OPR. Various methods are being investigated to achieve these goals, ranging from improved compressor three-dimensional blade designs to improved turbine cooling hole shapes and methods. Finally, a complementary effort in improving the accuracy, range, and speed of computational fluid mechanics (CFD) methods is proceeding to better capture the physical mechanisms underlying all these problems, for the purpose of improving understanding and future designs.